ML20043C362
| ML20043C362 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/31/1990 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20043C339 | List: |
| References | |
| NUDOCS 9006050103 | |
| Download: ML20043C362 (222) | |
Text
{{#Wiki_filter:Attachment til LICENSING REPORT i for l INCREASED STORAGE CAPACITY ! I for the JAMES A. FITZPATRICK PLANT l SPENT FUEL POOL l l l-New York Power Authority l L James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 DPR-59 ; b Of" pil.
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J 1' TABLE OF CONTENTS , SECTION DESCRIPTION g
1.0 INTRODUCTION
1-1 .. l 1.1 Introduction 1-1 2.0 MODULE LAYOUT FOR INCREASED STORAGE 2-1 2.1 New Proposed Racks - 2-1 1 2.2 Synopsis of Existing Racks in the Pool 2-2 l 3.0 RACK FABRICATION AND APPLICABLE CODES 3-1 3.1 Design Objective 3-1 ' 3.2 Anatomy of the Rack Module ' 3-2 I 3.3 Materials of Construction 3-6 3.4 Codes, Standards, and Practices 3-8 for the Spent Fuel Pool Modifioation 4.0 CRITICALITY SAFETY CONSIDERATIONS 4-1 , 4.1 Design Bases 4-1 4.2 Summary of criticality Safety Analyses 4-3 4.2.1 Normal Operating Conditions 4-3 4.2.2 Abnormal and Accident conditions 4-4 4.3 Reference Fuel Storage Cell 4-4 : 4.3.1 Fuel Assembly Design Specifications 4-4
' 4.3.2 Storage Rack Cell Specifications 4-4 .
1
TABLE OF CONTENTS ' (continued) ! 8 I EEC22QH DESCRIPTION 2333 4.4 Analytical Methodology 4-5 l 4.5 Criticality Analyses and Tolerance variations 4-6 4.5.1 Nominal Design Case 4-6 4.5.2 Uncertainties Due to Rack Manufacturing Tolerances 4-8 4.5.2.1 Boron Loading variation 4-6 . 4.5.2.2 Boral width Tolerance . variation 4-6 4.5.2.3 Storage cell Lattice Pitch variation 4-6 4.5.2.4 Stainless Steel Thickness Tolerances 4-7 ' 4.5.3.5 Siroonium Flow Channel Density variation 4-7 , 4.5.3 Reactivity Effects of Boral Asial Length 4-7 4.5.4 Water Gap Spacing between Modules 4-8 4.6 Abnormal and Accident Conditions 4-9 Temperature and Water Density 4.6.1 Effects 4-9 < 4.6.2 Abnormal Location of a , Fuel Assembly 4-9 l 4.6.3 Bosentric Fuel Assembly Positioning 4-10 4.6.4 Dropped Fuel Assembly 4-10 4.6.5, Fuel Rack Lateral Movement 4-10 4.7 References for Section 4 4-11 Appendia A to Section 4: Benchmark A-1 calenlations
y . . s TABLE OF CONTENTS y (continued) p. m SECTION DESCRIPTION g b 5.0 THERMt.L BYDRAULIC CONSIDERATIONS 5-1 0-I 5.1 Introduction 5-1
. 5.2 System Description 5-2 l 5.3 Decay Beat Load Calculations 5-6 ==
5.4- Mathematical Idealization of 5-7 the system E 5.5 Mathematical Model and Results 5-7 5.6 Time-to-Boil 5-10 g 5.7 Local Pool ~Nater Temperature 5.7.1 Basis 5-10 E 5-10 5.7.2 Model Description 5-11 7- 5.8 Cladding Temperature- 5-13 5.9 Blocked-Call Analysis 5-15 5.10 References 5-15 E
- 6.0 RACK STRUCTURAL CONSIDERATIONS n
U 6.1 Analysis outline L 6-1 6.2 Fuel Rack - Dynamic Model 6-4 [ 6.2.1 outline of Model for Computer Code DYNARACK 5- S 6.2.2 Model Description C .-8 6.2.3 Fluid Coupling 6-8 6.2.4 Damping 6-10 6.2.5 Impact 6-10 6.3 Assembly of the Dynamic Model 6-11
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TABLE OF CONTENTS (continued) SECTION ggscRIPTIQ3 pg 6.4 Time Integration of the Equations 6-14 of Motion 6.4.1 Tims Sistory Analp is Using 6 Multi-Degree of Freedom Rack Model 6.4.2 Evaluation of Potential for Inter-Rack Impact 6-16 6.5 Structural Acceptance Criteria 6-17 6.6 Material Properties 6-18 6.7 Stress Limits for Various conditions 6-19 6.7.1 Normal and Upset Conditions (Level A or Level 3) 6-19 6.7.2 Level D Service Limits 6-22 6.8 Results for the Analysis of Spent 6-22 Puel Racks Using a Single Rack Model and 3-D Seismic Motion 6.9 Impact Analyses 6-25 6.9.1 Impact L~ M ng Between Puel Assembly and Cell Wall 6-25
. 6.9.2 Impacts Between Adjacent Racks 6-25 6.10 Wald Stresses 6-26 6.10.1 Baseplate to Rack Welds and Cell-to-cell Welds 6-25 6.10.2 Beating of an Isolated cell 6-27 6.11 Seismic Qualification using Multiple 6-27 Time Bistories 6.12 Multi-Rack Analysis 6-28 f It'* h 9%R hO 4 *MP ='L5lth"' , q h gn 49 q g 9 $gp ig, G $+g .g 9 g g gggg y gg g.ig. g ,a e gg gg g g ,,g,
L p TABLE OF CONTENTS (continued) SECTION DESCRIPTION PAgg 6.13 Definition of Terms Used in Section 6.0 6-31 6.14 References 6-32 Appendix to Section 6 7.0 ACCIDENT ANALYSIS AND TIERMAL (SECONDARY) STRESSES 7-1 7.1 Introduction 7-1 7.2 Results of Accident Reevaluation 7-1 7.2.1 Fuel Pool 7-1 7.2.2 Fuel Storage Building 7-2 7.2.3 Refueling Accidents 7-2
\- 7.2.3.1 Dropped Fuel Assembly 7-2 7.2.3.2 Dropped-Gate 7-3 7.3 Local Buckling of Fuel call Walls 7-4 7.4 Analysis of Welded Joints in Rack 7-5 7.5 References 7-6 8.0 IW-8ERVICE SURVEILIANCE PROGRAM 8-1 8.1 Purpose 8-1 8.2 Coupon Surveillance 8-1 8.2.1 Description of Test coupons 8-1 8.2.2 Benctuaszk Data 8-2 8.2.3 Long Tara Surveillance 8-2 9.0 POOL STRUCTURAL ANALYS'Is 9-1 10.0 RADIOLOGICAL CONSIDERATIQRS lo-1 11.0 C08T/1Mnre1 PIT ASSESSMENT 10-1
)
1.0 INTRODUCTION
1.1 Introduction l James A. FitzPatrick (JAF) Nuclear Power Plant is a boiling water reactor (BWR) and is located on the southeast shore of Lake g ontario, approximately 6 miles northeast of the city of Oswego, How Ton. The plant le rated at 2436 Mwt and has been in commercial operation since July, 1975. l The spent fuel pool of the FitzPatrick plant was raracked in 1981 with " poisoned" high density racks made of aluminum alloy, increasing its storage capacity to 2244 locations. As indicated in Table 1.1 of this report, the current installed capacity will lead to loss of full core discharge capacity at the scheduled 1991 refueling. The projected loss of full core discharge capability l in 1991 prompted the New York Power Authority to undertake steps to increase spent fuel storage in the fuel pool. Fortunately, there is additional floor space available in the JAF spent fuel pool wherein supplemental modules can be installed. Under the proposed storage expansion, five modules containing a total of 553 storage locations will be added to the pool, increasing the total installed capacity to 2797 locations. As indicated by Table 1.1, the increased capacity will extend the date of loss-of-full core offload capability by sir years, to 1997. Similarly, the loss of normal batch offload capability will be extended to 2001, from the current projected date of 1995. The new spent fuel storage racks are free-standing and self supporting. The principal construction materials for the new l racks are ASTM A240-Type 304L stainless steel sheet and plate 1-1 1
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._,p.IuL.,._h stock, and A564-Type 630 (precipitation hardened stainless steel) for the- adjustable support spindles. The only. non-stainless material utilized in the racks is the neutron absorber material which is a composite of aluminum-boron carbide-sandwich available.
under the patented product BORALTM, The racks are designed and analyzed- following the rules of ASME Section III, Division 1, Sub-Section NF. The material-procurement and fabrication of . the rack modules are required to conform to 10 CFR 50 Appendix B requirements. CBI's Quality Assurance programmatic cossaitments as stated in the CBI Quality Assurance Manual conform to 10 CFR 50, Appendix B, and _are required to be met in all aspects of the project activity, including material procurement, fabrication, non-destructive
===4 amtion, vendor surveillance, qualification of test apparatus, design control, audits and record retention.
This Safety Analysis Report documents the design and analyses performed to demonstrate that the new spent fuel racks satisfy all governing requirements of the applicable - codes and standards as listed in section 3.4, in particular, 'OT Position for Review and Acceptance of Spent Fuel Storage and Bandling Applications", USNRC (1978) and the January 1979 revision thereto. The safety assessment of the proposed rack modules invcived demonstration of their hydrothermal, criticality and structural adequacy. Bydrothermal adequacy requires that fuel eladding will not fail due to excessive thermal stress, and, that the steady state pool bulk temperature will remain low such that the reinforced concrete wall and slab are not overstressed and that
- the steady state temperatures conform to ACI 349 guidelines.
1-2 l
9 Demonstration of structural adequacy primarily involves the proof that the freestanding modules will 'not collide under the postulated Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (OBE) events, and that the primary stresses in the module structure will remain below the ASME Code allowablea. Finally, the structural qualification also includes analytical proof to demonstrate that the 'sub-criticality ' of the stored fuel will be maintained under accident scenarios such as fuel assembly drop. Criticality safety Analysis shows that the effective neutron multiplication factor (k gg) for the stored fuel array is bounded by the USNRC limit of 0.95 'under assumptions of 95% probability and 95% confidence. Consequences of inadvertent placement of fuel assembly are also evaluated as an essential aspect of criticality analysis. The criticality analysis sets the requirements > on the length of.the B-10 acreen and the areal B-10 density. The following sections in this report contain a concise and systematic documentation of the analyses performed to demonstrate the large margins of safety with respect to all USNRC specified criteria. In sumary, exhaustive analyses have shown the racks and spent fuel pool system design exceed the following criteria.
- 1. The offactive multiplication factor (kegg) of less than 0.95 is maintained for all possible operating and accident conditions.
- 2. Adequate cooling under both normal and abnormal fuel unloading rates is maintained, and special operating conditions are defined in the event of loss of coolant.
1-3
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- 4
- 3. The racks and pool structure will remain functional and !
withstand earthquake loadings ard any probable accident conditions.
- 4. Radiological doses are within acceptable limits. I t
t I, i-S f 1-4 - - - - - - - - - - . . _ _ _ _ _ a -
t a Table 1.1 FUEL DISCHARGE DATA-EXCESS EXCESS TOTAL STORAGE STORAGE NUMBEM OF NUMBER OF WITBOUT AFTER
*?ISCHARGE BATCH ASSEMBLIES ASSEMBLIES RACK RACK DATE No. DISCBARGED STORED ADDITION ADDITION 6/1977 1 132 132 628 628 9/1978 2 136 268 492 492 5/1980 3 160 428 332 332 11/1981* 4 188 616 1628 1628 6/1983 5 200 816 1428 1428 2/1985 6 196 1012 1232 1232 1/1987 7 188 1200 1044 1044 8/1988 8 184 1384 860 860 3/1990 9 156 1540 704 1257 10/1991 10 208 1748 496* 1049 10/1993 11 204 1952 292 845 10/1995 12 208 2160 84** 637 10/1997 13 208 2368 .------ 429*
10/1999 14 208 2576 ------ 221 10/2001 15 208 2784 ------ 13** Loss of full core offload capability. Loss of normal batch offlos4 capability. 1-5
. . . . - . . . ~ . . .. . . . . .. .. ..
____.._7_._ . . . . . . . a 2.0 MODULE LAYOUT FOR INCREASED STORAGE 2.1 New Pronosed Racks The James A. FitzPatrick high density spent fuel storage racks consist of individual cells with 6.10" (min.) inside square dimension, each of which accommodates a single Boiling Water Reactor (BWR) fuel assembly. The fuel assembly can'be stored in the storage locations in channelled or unchannelled configuration. Table 2.1 gives the essential storage cell design data. The rack modules proposed to be emplaced in the FitzPatrick pool are in three discrete sizes, denoted as Modules A, B and C, respectively. Table 2.2 gives the pool module data. Thus, altogether there are 553 added storage locations in-the fuel pool. Figure 2.1 shows the module layout for the added storage . The existing and new modules for the FitzPatrick fuel pool are qualified as non-impacting freestanding racks, i.e., each module is freestanding and is shown to undergo minimal kinematic displacements during the postulated seismic events. Thus, rack-to-rack or rack-to-wall impacts are precluded. The rack module support legs are of remotely adjustable type. Figure 2.2 shows a typical new rack module for the FitzPatrick fuel pool. 2-1
. 3, 9
L 2.2 Svnensis of Tristina Racks in the Pool Like the new proposed racks, the existing racks are full length, top entry type, designed to maintain spent fuel assemblies in a space geometry which precludes the possibility of criticality under normal and abnormal conditions. Normal conditions exist when the spent fuel assemblies are stored in the spent fuel storage racks in the design storage position. Abnormal conditions may result through equipment mishandling or from rack deflections due to earthquake loadings. The fully loaded existing spent fuel storage racks are designed to seismic class I requirements per the FitzPatrick plant Final Safety Analysis Report (FSAR). The existing fuel storage racks contain a storage capacity of 2244 fuel assemblies. The fuel assemblies are stored in anodized aluminua modules. The modules are interconnected in a group to
=lai=4se relative displacement and to prevent impact. In order to optimize storage space the modules are arranged in arrays of 8x10, 8x8, or .11.x10 (as shown on Fig. 2.3). The fuel assemblies are inserted into cavities that are formed by a cluster of cans.that are arranged in a checkerboard pattern (as shown on Fig. 2.4 ) .
The can provides seperation and lateral restraint for each fuel assembly Boral is sealed in cavities within each can by welding. A structural detail of four cans is shown on Fig. 2.5. The cans are constrained by upper and lower castings that are bolted to plates along the perimeter to form a box structure (see Fig. 2.6). The lower castlng vertically supports each fuel assembly. - Each-module is free-standing with no lateral restraints to the wall and is supported by four steel feet that transfer load to the pool floor. The lateral loads on the racks are transferred by friction between the feet and the pool floor liner plate. 2-2 l
The existing spent fuel storage racks are made up of double-walled aluminum containers (as shown on Fig. 2.7). These are approximately 14 feet long and have a square cross section with an inner dimension-of 6.16 inches. The nominal pitch between fuel assemblies is 6.625 inches. With a fuel channel loaded onto a fuel assembly, the maximum square dimension is 5.768 inches. Without the channel, the maximum square dimension of the assembly is 5.470 inches (at the lower tie ~ plate). Therefore, no interference problems are found in loading spent fuel into the existing racks. A Boral plate is seal welded in the cavity between adjacent fuel assemblies. The minimum amount of boron-ten (Bl o) per unit area of Boral plate is 0.0232 grams per square centimeter. This is equivalent to 1.4 x 10 21 boron-ten atoms per square centimeter. To meet seismic Class ! requirements, the storage racks are designed so that stresses in a fully loaded' rack, subject to specified ' earthquake loadings, do not exceed allowable stresses recommended by the American Society of Civil Engineers (ASCE) Task Force Committee on lightweight alloys. For areas within the' rack where stresses are complex and dif ficult to analyze, structural design is based on results of load tests. In addition, the storage racks are designed so that permanent distortion et the structure does not occur under application of forces equal to the capacity of the fuel' handling hoists. NRC Letter from T.A. Ippolito to G.P. Berry dated June 18, 1981 transmitted approval for use of the existing racks. 2-3
g i l Table 2.1 f Design Data for New Racks
' I I.D.
6.10 inch (min.) (inside dimension) Cell Minimum Pitch (.295 inch .i cell N=4 a=1 Pitch 6.355 inch Boral Loading (min.) .0135 ga per sq.ca. (B-10): Boral plate width: 5 inch.t. 06 inch Boral picture frame (bounding) size- 5.125" x 144-1/2* Boral minien= length: 144 inches cell height: 170 inch , Baseplate thickness: 1/2 inch Bottom plenua-height: 11.63 inch (naminal) Number of supports per modules Four (miniun=) Support Type Remotely adjustable 2-4 1
1 l- , Table 2.2 n. l MODULE DATA. If , 4 I' Number of Calla Per Modula Module No. of Dry Module I.D. N-8 E-W Total Modules Weight (lb) J l-(Fig. 2.1) A 11 11 121 One 12500 ! B 12 11 132 Two 13900 C 14 6 84 Two 8850
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e FIGURE 2.6 STRUCTURAL DETAIL - PARTIAL EI.EVATION (EXISTING RACKS) 2-11
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f i i 3.0 RACK FABRICATION AND APPLICABLE CODES ' 1 3.1 Damien obinetive The central objective governing the design of the new high density _ storage racks for ' the James A. FitzPatrick fuel' pool is defined in the following six criteria ' i (1) The rack module is fabricated in such a manner that there is na weld splatter on the storage cell surfaces L which would come in contact with the fuel assembly. Waldl l splatter on the lateral surface of the storage cell, which can come in contact with fuel assemblies, can be l detrimental to its structural integrity. (ii) The storage locations are designed and constructed in- -4 h such a way that redundant flow paths for the-coolant are I available in case the main designated flow path is l-blocked. (iii) The fabrication process based on the rack design involves ' ope' rational sequences which' permit immediate L. and convenient verification by the inspection staff to ensure that the " poison
- panels-are correctly installed..
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l i (iv) The storage cells. are connected to each other by autogenously produced corner welds which leads to a honeycomb lattice construction. The extent of-welding is selected to "detune" the racks from the ground motion (SSE or OBE), such that the rack . displacements are minimized. (v) The baseplate provides a conformal contact surface for the a nose" of the fuel assembly. (vi) The module design affords built-in flexibility in the fabrication process so as to maintain the desired cell pitch even if certain aboxes" are slightly oversize. The foregoing objectives are - fully realised in the -module design presented in this section. 3.2 hatomv of +ha eaak N~4nia The new high density rack module design employs storage cell locations with a single poison panel sandwiched between adjacent austenitic steel surfaces. A complete description of the rack geometry is best presented by first introducing its constituent parts. The principal parts are denoted ass (1) the storage box subassembly (2) the baseplate (3) the neutron absorber material, (4) picture frame sheathing, and (5) support legs. Each part is briefly described below with the aid of sketches. 3-2 I l
(1) Storage cell box subassembly: The so-called " boxes" are fabricated from two precision formed channels- by seam welding them together in a seam welding machine equipped with copper chill bars, and pneumatic clamps to minimize distortion due to welding-heat input. Figure 3.1 shows the " box". The minimum weld penetration is 80% of the box metal - gage- which is 0.075" (14 gage). The boxes are manufactured to 6.10" minimum I.D. (inside dimension). As shown in Figure 3.1, each box has two lateral holes punched near its bottom edge to provide auxiliary -flow holes. In the next step, a picture frame sheathing is press formed in a precision die. The sheathing is shown attached to the box in Figure 3.2. The sheathing - is made to a precise offset of .085 2 .005" to ensure an unconstrained installation of Boral plates. The." picture frame sheathing" is attached to each side of the box with the poison material (Boral) installed 'in- the sheathing cavity. The top of the sheathing is connected using a smooth continuous fillet weld near the top of the box. The edges of the sheathing . and' the box are welded together to form a smooth lead-in edge. The box with integrally connected sheathing is referred to as the " composite box". 3-3
- 1 The " composite boxes" are arranged in a checkerbcard array to ferm an assemblage of storage cell locations-(Figure 3.3). 'The inter-box welding and- pitch adjustment is accomplished by small longitudinal ;
austenitic stainless connectors shown as.small circles-in Figure 3.3. An assemblage of box assemblies thus prepared is welded edge to edge as shown in Figure 3.3 resulting in a , honeycomb structure with axial, flexural and torsional rigidity depending on the extent of intercell welding - , provided. Figure 3.3 shows that two edges of each interior box are connected to the contiguous boxes resulting in a well defined path for " shear flow", - (2) Baseplate: The baseplate, 1/2 inch thick, provides a centinuous hori: ental surface fer supperting the fuel assemblies. The baseplate has a concentric hele with a 450 taper in each cell location to provide a seating surface conforming to the fuel assembly. The baseplate is attached to the cell assemblage . by fillet welds. (3) The neutron abscrber material: As centioned earlier, Boral is used as the neutron absorber material. Boral is manufactured by AAR Erecks and Perkins of Livenia, L Michigan, More en this material follows in the next y section. 1' l l l l l 3-4 l
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L> , 1 l l l! l L l L l L (4) Picture Frame Sheathing: The sheathing is a part of the l
" composite box assembly" described earlier. The
); sheathing serves as the locater and retainer of the
-poison material. As such, -it is .made in repeatable precise dimensions. This is accomplished by press-I 1=
forming stainless sheet stock in a specially high ! tolerance die. The schematic of the sheathing.is shown f in Figure 3.2. Figure 3.4 shows three storage cells in elevation with the fuel assembly shown in phantom in. one cell. The poison screen extends over 144" minimum vertical distance, straddling the active fuel length of 150 inches with the latter extending 3 inches in each end, beyond the poison screen. (5) Support Legs: As stated earlier, all support lege are the adjustable type (Figure 3.5). The top '(female) , position is made of austenitic steel material. The bottom part is made of 17:4 Ph series stainless steel to avoid galling. The support leg is equipped with a socket to - enable r remote leveling of the rack after its placement in the pool. Since these rack modules az. designed to be non-impacting, they are not equipped with " girdle - bars" to withstand inter-rack or rack-to-wall impacts. 3-5
'i i
The exterior surfaces of modules, where an inadvertent 1 placement of fuel on the outside of the module can ! occur, are equipped with " stand-of f" angles to prevent the misplaced fuel assembly from violating the ' required inter-assembly gap with respect to the other stored fuel assemblies. 3.3 MATERIALS OF CONSTRUCTION The principal material of construction utill ed in the fabrication- of the FitzPatrick plant' high density. racks is austenitic stainless steel (ASTM 240-304L). One notable exception l is the support spindle material which' is made out of a special high strength (precipitation hardened) stainless steel (A564-630). In addition to the structural and non-structural stainless material, the racks employ BoralTM, a patented product of _ AAR - Brooks and Perkins, as the neutron absorber material. A brief ; description of Boral and a list of fuel pools in which it is used follows. 1 Boral is a thermal neutron poison material camposed of boron p carbide and 1100 alloy aluminum. Baron carbide is a compound having a high boron content in'a physically stable and chemically inert fora. The 1100 alloy aluminum is a light-weight metal with high tensile strength which is p:cotected from corrosion by a highly resistant oxide film. The two materials, boron carbide and ' aluminum, are chemically compe.cible and ideally suited for long-term use in the radiation, th==1 and chemical environment of the. spent fuel pool. - 3-6
Boral's use in. the spent fuel pools as a preferred neutron absorbing material can be attributed to the following reasons: (i) The content and placement cf boron carbide provides a very high removal cross section for thermal neutrons. (ii) -Boron carbide, in the form of fine particles, is homogenously dispersed throughout the central layer. of the Boral panels. (iii) The boron carbide and aluminum materials in Boral are not detrimentally affected by long-term exposure to gasuna radiation. (iv) The neutron absorbing central layer of Boral is clad with permanently bonded surfaces of aluminum. (v) Boral is stable, strong, durable,- and corrosion resistant. Boral is manufactured.under the control and surveillance of a Quality Assurance / Quality Control Program that conforms to the requirements of 10 CFR 50 Appendix B, " Quality Assurance Criteria for Nuclear Power Plants % As. indicated in Table 3.1, Boral has been licensed by the USNRC for use in numerous BWR and PWR spent fuel storage racks. Boral Material characteristics Aluminum: 1100 ailoy aluminum is the metallic ingredient of Boral. The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops'on its surface from exposure to the atmosphere or water. This film-prevents the loss of metal from general corrosion or pitting corrosion and the film remaina stable between a pH range of 4.5 to 8.5. 3-7
i Boron Carbide: The boron carbide contained in Boral is a , fine granulated powder that conforms to ASTM C-750-80 nuclear
; grade Type III. The particles range in size between 60 and 200 '
~
mesh, and the material conforms to the chemical composition and-properties listed in Table 3.2. ( A large body of test data and plant operating experience data l is available in . publications in the.public domain by Boral's manufacturer. i' 1, L 3;4 cones, svinn e n. Ann PuicTIcrE FOR THE SPENT FUEL POOL .i! L NODIFICATION The fabrication of the rack modules is performed . under a , strict. quality assurance program which meets 10 CFR 50 Appendix B 4 l- requirements. ! l The following codes, standards and practices were used for l all applicable aspects of the design, construction, and assembly of the spent fuel storage racks. Additional specific references ,
- related to detailed analyses are given in each section. '
Design codes a. (1) AISC Manual of Steel Construction, 8th . Edition, 1980 (provides detailed structural criteria for-linear type supports). ANSI N210-1976, ' Design Objectivss for Light Water (2) Reactor Spent Fuel Storage Facilities at Nuclear Power Stations * (contains> guidelines for fuel rack design). t 3-8 i w ,, , , , , - ,n ,,
(3) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section III, 1983 ' Edition up to and including Summer 1983 Addenda (Subsection NF) (governing stracture.1 design code). r (4) It d, 1986 Edition, including up to 1988 Addenda (governing material procurement, fabrication and , NDE). (5) ASNT-TC-1A June, 1980 American Society for Mondestructive Testing (Recosumended Practice for Personnel Qualifications).
- h. neatselmi cad-= - standmeda of asThan (1) E165 - Standard Methods for Liquid Penetrant ;
Inspection (2) A240 - Standard Specification for Beat-Resisting Chromium and Chromium-Nickel stainless steel Plate, Sheet and Strip for Fusion-Welded Unfired Pressure vessels , (3) A262 - Detecting susceptibility to Intergranular Attack in Austenitic Stainlese steel ! (4) A276 - Standard Specification for stainless and Beat-Resisting Steel Bars and Shapes (5) A479 - Steel Bars for Boilers & Pressure vessels (6) C750 - Standard Specification for Nuclear-Grade Boron Carbide Powder (7) C992 - Standard Specification for Baron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Puel Storage Racks (8) American Society of Nechanical Engineers (ASME), i Boiler and Pressure vessel Code, Section II-Parts A l and C, 1986 Edition, up to and including 1988 Addenda. ; 4 3-9
.- - . - - - . . - - . - . .- . . - . - . . ~ - . . - . - - . - - - . . - . . - _ - .
. J I
i I
- c. Walding codes l
ASME Soiler and Pressure Vessel Code, Section IX-Welding and Srasing Qualifications, 1986 Edition up to j and including 1988 Addenda. l
- d. Quality Amanranea, clean 11naar. packanina, shineine.
Raemivina. Starmaa. and sandlina manniramanta (1) ANSI M45.2.2 - Packaging, Shipping, Receiving, Storage and Sandling of Items for Nuclear Power . Plants. * (2) ANSI 45.2.1 - Cleaning of Fluid Systems and Associated Components during Construction Phase of ' Nuclear Power Plants. j (3) ASME Boiler and Pressure vessel, Section V, Wondestructive Examination, 1983 Edition, including Susener and Ninter 1983. - l (4) ANSI - N45.2.11, 1974 Quality Assurance Requirements for the Design of Nuclear Power Plants. (5) ANSI - N45.2.5 - Qualifications of Inspection, armaination, and Testing Personnel for Nuclear Power Plants (Regulatory Guide 1.50). , (6) ANSI N45.2.13 - Quality Assurance Requirements for i control of Procurement of Equipment Materials rad , Services for Nnclear Power Planta (Regulatch:y Guide . 1.123). t (7) ANSI R45.2.23 - Qualificat; ion of Quality Assurance Program Audit Personas), for Nuclear Power Plants - l (Regulatory Guide 1.146), ,
- e. Gerarning M Danign Dnenannts '
(1) NUREG-0800, Standard Review Pica (1981). , r 3-10 :
i .
]
l l l 4 (2) *0T Position for Review and Acceptance of spent Fuel Storage and Randling Applications," dated j April 14, 1978, and the modifications to this j document of January 18, 1979. . u
- f. Other ANSI Standards (not listed in the preceding)
(1) N45.2 - Quality Assurance Program Requirements for i Nuclear Facilities - 1971 .i (2) N45.2.2 - Packaging, Shipping, Receiving, Storage [ and sandling of Items for Nuclear Power Plants (during the Construction phase) - 1972 : (3) N45.2.9 - Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants - 1974 ; t (4) N45.2.10 - Quality Assurance Terms and Definitions '
- 1973 l
(5) N210 - Design Objective for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants t
- g. Cada-af-yaderal Reenlatiana ,
(1) 10 CFR 21 - Neporting of Defects and Non-compliance ! (2) 10 CFR 50 - Appendix A - General Design Criteria for Nuclear Power Plants (3) 10 CFR 50 - Appendix 3 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
- h. Regulatory anidas (1) RG 1.13 - Spent Fuel Storage Facility Design Basis (2) RG 1.25 - Ass tions Used for Evaluating the Potential Radiol ioal Consequences of a Fuel Bandling Accident the Fuel Bandling and Storage -
g Facility of Boiling and Pressurised Water Reactors. 'q_ 0 7 I 3-11
' ' ~ ' ' " ' '
l 4 l i (3) RG 1.28 - (endorses ANSI N45.2) - Quality Assurance l Program Requirements, June,1972. 1 (4) RG 1.29 - Seismic Design ClassificatiQn (5) RG 1.38 - (endorses ANSI N45.2.2) Qualit ! Requirements for Packaging, Shipping, y Receiving, Assurance j storage and Handling of Items for Water-cooled Nuclear Power Plants, March, 1973. (6) RG 1.44 - Control of the Use of Sensitised stainless steel (7) RG 1.58 - (endorses ANSI N45.2.6) Qualification of ! Nuclear Power Plant Inspection, Examination, and l , Testing Personnel. Rev. 1, September, 1980 1 1.64 (8) RG - (endorses ANSI N45.2.11) Quality Assurance Requirements for the Design of Nuclear
- Power Plants, October, 1973.
1.74 (9) RG - (endorses ANSI N45.2.10) Quality Assurance Terms and Definitions, February, 1974. (10) RG 1.88 - (endorses ANSI N45.2.9) Collection, i Storage and Maintenance of Nuclear Power Plant ! l Quality Assurance Records. Rev. 2, October, 1976. , (11) Rs 1.92 - Combining Modal Responses and spatial components in seismic Response Analysis (12) RG 1.123 - (endorses - ANSI N45.2.13) Quality i' Assurance Requirements for control of Procurement of Items and Services for Nuclear Power Plants. .
- i. Branch Technical Desition (1) CPB 9.1 Criticality in Puel storage Facilities j (2) Asa 9-2 - Residual Decay Energy for Light-Water Reactors for Long-Term Cooling k
l r 3-12 l h 1
t 6 i
- j. Standard Review Plan (1) SRP 3.7.1 - seismic Design Parameters (2) SRP 3.7.2 - Seismic System Analysis j (3) SRP 3.7.2 - Seismic subsystem Analysis
]
(4) SRP 3.W.4 - Dther Seismic Category I Structures (including Appendix D) , i (5) SRP 9.1.2 - Spent Fuel Storage < (6) SRP 9.1.3 - Spent Fuel Pool Cooling and cleanup Systes ,
- k. Q1 hat James A. Fit Fatrick Final Safety Analysis Report (FSAR)- ,
e f e 4 3-13
--ny
;~ - .
l l
. i l
Table 3.1 i Boral Experience List (Domestic and Foreign) { Pressurised Water Roastors I Vented , construc- Mfg. i Plant Utility tion Year j . Bellefont 1, 2 Tennessee Valley Authority No 1981 1 D.C. Cook 1,2 Indiana & Michigan Electr; c No 1979 , Indian Point 3 NY Power Authority Yes 1987 Maine Yankee Maine Yankee Atom;.c Power .Yes 1977 3 1 salen 1, 2 Public service Elec & Gas No 1980 Seabrook New Bampshire Yankee No --- Sequoyah 1,2 Tennessee Valley Authority No 1979 Yankee Rowe Yankee Atomic Power Yes 1964/1983 Eion 1,2 Casusonwealth Edison Co. Yes 1980 l Byron 1,2 Casusonwealth Edison Co. Yes 1988 Braidwood 1,2 j Commonwealth Edison Co. Yes 1988 - Yankee Rowe Yankee. Atomic Electric Yes 1988 ' i h814== Water Reestors Browns Perry 1,2(3 Tennessee Valley Authority Yes 1980 Brunswick 1,2 Carolina Power & Light Yes 1981 , Clinton Illinois Power Yes 1981 g Cooper Nebraska Public Power Yes 1979 , Dresden 2,3 Commonwealth Edison Co.- Yes 1981 Duane Arnold Iowa Elec. Light & Power No 1979 l J.A. FitsPatrick NY Power Authority No 1978 E.I. Eatch 1,2 Georyia Power Yes 1981 Bope Creek Publ; c Service Elec & Gas Yes 1985 Eumboldt Bay Psicific Gas & Electric Yes 1986 Lacrosse Dairyland Power Yes 1976 Limerick 1,2 Philadelphia Electric No 1980 Monticello Northern States Power Yes 1978 Peachbotton 2,3 Philadelphia Electric No 1980 Perry, 1,2 Cleveland Elec. Illuminating No 1979 Filgria Boston Edison No 1978 Shoreham Long Island Lighting Yes --- - Susquehanna 1,2 Pennsylvania Power & Light No 1979 Vermont Yankee Vermont Yankee Atomic Power Yes 1978/1986
. Hope Creek Public Service Elec & Gas Yes 1989 3-14 -
--y--ww--wrw:-r uu' r te ner "w-r '
i l Table 3.1 (continued) l Foreiga zastallations seing moral l Franoe i i
, 12 PWR Plaats Electricite de France l i
- SOE%I AfriSa Koeberg 1,2 ESCON Switserland ;
i ' Besnau 1,2 Nordostschweizerische Kraftwerke AG Googen Kernkraftwerk Gosgen-Dan nan AG ; ! I Taiwas Chin-Shan 1,2 Taiwan Power Company Kuosheng 1,2 Taiwan Power Company
?
1 l 3-15
l l 1 Table 3.2 Boron carbide chemical composition, Weight %* l Total boron 70.0 min. i 310 isotopic content in 18.0 natural baron Boric oxide 3.0 max. , Iron 2.0 max. Total boron plus 94.0 min. total carbon Boron carbide Physical Properties chemical formula 3c 4 , i Boron content (weight) 78.28% carbon content (weight) 21.724 l Crystal structure rombobedral ' l Density 2.51 gm./oc-0.0907 lb/en. in. Neiting Point 24500 C - 4442 0 y Boiling Point . 35000 C-6332 0 y Microscopic capture 600 barn cross section Provided by AAR Brooks & Perkins 3-16 l
. . _ _ _ - . . .. .__'._..____.... - ._ _ .._. _ _ _ _ _ . _ ,_. _ . __ _ _ ,.~ _ _ _ .. _ _ .
i i i i 1 1 I t . i 1 FLOW HOLI , e , t s i MM O , t FIGURE 3.1 SEAM WELDING PRECISION FORMED CHANNELS 9 4 3-17 4 4 J
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i I i -- ' l muummummum M FIGURE 3.3 A CROSS SECTIONAL VIEW 0F AN ARRAY OF STORAGE LOCATIONS 3-19
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,.. , ... . . . . . > . . . . . . . . . ~ ~ - * ' - * ~ ' ' ' * ' " * * * * " ' ~ ~ '~'
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(4) 1" dia. holes , 9 6" SQUARE : e FIGURE 3.5 ADJUSTABLE SUPPORT - 4 e 4 3-21 _ _ _ . - . - , . - * **^
4.0 CRITICALITY SAFETY ANALYSES 4.1 DESIGN BASES The high density spent fuel storage racks for the J. A. FitzPatrick Nuclear Power Plant are designed to assure that the ; neutzen multiplication f actor (k gg) is less than 0.95 with the ; racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with non-borated water at a temperature corresponding to the highest reactivity. The design basis fuel for the storage rack is an 8x8 BWR fuel rod assembly with a uniform enrichment of 3.3 wtt U-235. The maximum calculated reactivity of the storage rack includes a margin for uncertainty in reactivity calculations and in mechanical tolerances, statisti-l cally combined, such that the true k gg will be less than 0.95 L with a 954 probability at a 954 confidence level. Reactivity l effects of abnormal and accident conditions have also been evaluated to assure that under credible abnormal conditions, the reactivity will be less than the limiting design basis value. Applicable codes, standards, and regulations, or pertinent sections thereof, include the following: O General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling. O USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3, July 1981 O USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including the modification letter dated January 18, 1979. USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility l 0 Design Basis, Rev. 2 (proposed), December, 1981. i 4-1
1 l l. I O USNRC Regulatog Guide 3.41, validation of Calculational i Methods for Nuclear Criticality Safety (and related ANSI l N16.9-1975). 2 O ANSI /ANS-57.2-1983, Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power q Plants. ! O ANSI N210-1976, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants. O ANS-8.17-1984, Criticality Safety criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors. O ANSI N16.1-1975, Nuclear Criticality Safety in Opera-tions with Fissionable Materials Outside Reactors. To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were , l made: l 0 The racks are assumed to contain the most reactive fuel authorized to be stored in the facility. 7 O Moderater is pure, unborated water at a temperature L within the design basis range corresponding to the t l highest reactivity.
- O Criticality safety analyses are based upon k., i.e.,
- lattice of storage racks is assumed infinite in all
- directions. No credit is taken for axial or radial '
neutron leakage (except as necessary in the assessment l of abnormal / accident conditions). i l O Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water. O The Boron-10 in the Boral absorber plates were assumed to be uniform with a minimum areal density corresponding to the lower limit of the manufacturing tolerance. , 4-2 l
l l 4.2
SUMMARY
OF CRITICALITY SAFETY ANALYSES , I 4.2.1 Normal Ocaratino Conditlens The basic calculations supporting the criticality safety : of the Fit: Patrick fuel storage racks are sununarized in Table 4.1. F Based upon a uniform enrichment of 3.3% U-235, the maximum b in i the storage rack is 0.937* (954 probability at the 95% confidence ! level) including all known uncertainties. Thus, the fuel storage rack satisfies the design basis requirement of a maximum k.gg less I than or equal to 0.95. i The design basis criterion of a uniform enrichment of 3.3% without gadolinium burnable poison is very conservative. . With the gadolinium burnable poison loading usually contained in BWR fuel, the actual reactivity will be considerably lower, even at the peak reactivity over burnup, where the gadolinium is essen-tially consumed. For this reason, the racks can safely accom-modate fuel of higher enrichments with the gadolinium burnable poison normally contained in BWR fuel. 4.2.2 Abnormal and Accident Conditions None of the credible abnormal or accident conditions that have been identified will result in exceeding the limiting reactivity (k fg of 0.95). The effects on reactivity of credible abnormal and accident conditions are summarized in Table 4.2. No other credible accident events or abnormal configurations have been identified that might have any adverse effect on the storage e rack criticality safety. The double contingency principle specifically invoked in the definitive USNRC letter of April 14, 1978 precludes the necessity of considering the occurrence of
- b is calculated for an infinite array, neglecting neutron ,
leakage. 4-3
.-. - . . . . - = -- - . . . . .. . -
1 I l more than a single unlikely and independent accident condition I concurrently.
)
i 4.3 REFERENCE FUEL STORAGE CELL
]
I 4.3.1 Fuel Assembly Desian Snecification The design basis fuel assembly is a standard 8x8 array of BWR fuel rods containing 002 clad in Eircaloy. Design parame-ters are sunanarized in Table 4.3. With the-design basis uniform enrichment of 3.30 wtt U-235, this fuel assembly has a minimum k. of 1.368 (954/954) in the uncontrolled reactor lattice geometry at l 68'r, corrected for bias and uncertainties. 4.3.2 Storace Rack Call Snacifications l l The design basis storage rack cell consists of-an egg-crate structure, illustrated in Figure 4.1, with fixed neutron l absorber material (Boral) of 0.0146 g/cm2 boron-10 areal density (0.0135 g B-10/cm2 minimum) positioned between the fuel assembly storage cells in a 0.085 inch wide space. This arrangement provides a nominal center-to-center lattice spacing of 6.355 inches. Manufacturing tolerances, used in evaluating uncertain- , ties in reactivity, are indicated in Figure 4.1. The 0.075-in. stainless-steel box which defines the fuel assembly storage cell has a nominal inside dimension of 6.16 in. This allows adequate clearance for inserting or removing the fuel assemblies, with or - without the Zircaloy channel. Boral panels are not needed or used on the exterior walls of modules facing non-fueled regions, i.e., the pool walls. A 2 inch water gap (1h inch minimum) along the interface between modules would normally eliminate the need for poison panels on these module walls. However, as protection against hypothetical module movement during a seismic event, Boral y panels are used on the walls of alternate cells along one side of ! the interface. 4-4 l l i
l I i i l 1 4.4 ANALYTICAL NETHODOLOGY I l Criticality analyses of the high density spent fuel l storage racks were performed using the CASHC- t codetal, a two- i diment! %al multi-group transport theory code. Independent I verification calculations were made with the AHFX-MENO computer L packagetas, using the 27-group SCALE
- eross-section libraryt33 ;
with the NITAWL subroutine for U-234 resonance shielding effects (Nordheim integral treatment). Benchmark calculations are i presented in Appendix A and indicate a bias of 0.0013 a 0.0018 for CASNo-22 and 0.0106 a 0.0048 (95%/95%) for NITAWL-KENO. These methods of analysis and the benchmarking calculations have l previously been used in the evaluation of spent fuel storage l racks that have been reviewed and accepted by the USNRC. 9 In the geometric model used in the calculations, each' (. fuel rod and its cladding were described explicitly and reflect-ing boundary conditions (aero neutron currentl were used in the axial direction and at the centerline of. the Boral and steel ; plates between storage cells. These boundary conditions have the effect of creating an infinite array of storage cells in all directions. The CASM0-25 computer code was used as the primary method of analysis as well as a means of evaluating small reactivity increments associated with manufacturing tolerances. Diffusion theory constants, edited in the output of CASH 0-25, were used in the PD07 codete> or in a one dimensional diffusion theory routine to evaluate reactivity effects of the- Boral panel ' length, of the water gap between storage modules and the conse-quences of postulated accidents. SCALE is an acronym for Etandardized Eomputer &nalysis for kicensing [ valuation, a standard cross-section set developed by ORNL for the USNRC. 4-5 P
4.5 CRITICALITY ANALYSB8 AND TOLBRANCB VARIATICBS 4.5.1 Banimal Damien casa for the design basis reactivity calculattens with a unifera enrisbaent el 3.3%, the mestaal storage sell intimate multiplicaties faster, k. , is 0.9244 (bias correet&d CA8N0). With a serrection of +0.0061 4k for the statsua 8-10 leading and a ok of 0.0071 for all known uncertataties statistically coa-hiaed, the maataea k. in the fuel rack is 0.937 vhich is less than the desiga basis limit of 0.95 for k se. Independent sheek ealestations with AMPI-EBBC gave a k. of 0.924 a 0.008 (95%/95%, serrested for bias and temperature) whieb is la reassamble agreement with the reference CA8MO salculation. The R-faster for 95% probability at a 98% esafidence level was determined trea NB8 Bandbook 91:ss. 4.5.2 namarentatfaa ama to m*=k Manufaaturine valerammes 4.5.2.1 Baram Landiae Variatina The Beral absorber panels used in the storage sells are nestaally 0.075-inch thiek, with a 3-10 areal density of 0.0146 g/ema. The manufasturing toleranee 11att is 80.0011 g/eas in B-10 content. This assures that the staines borea-10 areal density at any location will act be less than 0.0138 gram /ema. Althoech this toleramee is statistically independent, for addittomal commerwatisa the referense eriticality salsulation assumed the statoes 8-10 loading (0.0135 g/sa s) everywhere. 4.5.2.2 marai Width Talaranea variation The referenee storage eell desiga esos a Seral panel
~
width of 5.00 a 0.04 taebes. For a redesties in width of the aasiana tolerance, 0.04 Lash, the saleslated positive reactivity toerement is +0.0017 ok. 4-4
)
i 4.5.2.3 storace cell Lattice Pitch variation l The design storage cell lattice spacing between fuel assemblies is 6.35! inch. Increasing the lattice pitch reduces 1 reactivity. For the manufacturing tolerance of 20.06 in., the corresponding maximum uncertainty in reactivity is t 0.0048 ok as determined by differential CASHO calculations. 4.5.2.4 stainless Steel Thiekness Tolerances The nominal thickness of the stainless steel box is 0.075 t 0.007 inch and 0.035
- 0.003 inch for the steel backing plate. The maximum positive reactivity effect of the expected stainless steel thickness tolerances was calculated (CASH 0-2E) to be 2 0.0004 ok.
4.5.2.5 "ireenium Flow Channel The design basis calculations assumed a flow channel thickness of 0.080 inches. Elimination of the :1rconium flow channel results in a small (- 0.006 ok) decrease in reactivity. 4.E.3 Reactivity Effects of Boral Axial Leneth Because of neutron leakage from the ends of the fuel assemblies in storage, Boral at the ends would serve little or no purpose and,therefore,is not necessary. Using diffusion theory constants edited in the CASHO-2E output- (reference design with and without Boral), one-dimensional axial calculations were made to evaluate the reactivity effect of reduced Boral axial lengths (cutback). These calculations used a thick (30 cm.) water re f 2.e ct: and neglected the higher absorptien of the stainless steel structural material above and below the active fuel. With fresh unburned fuel, and without credit for the distribution in enrichments or the gadolinium burnable poison actually present, a 4 -7
r i I Boral length of 3 inches less than the enriched fusi sone, top i and bettes, was f ound to be asseptable, i.e. , the k.s e with axial >
. leakage included was less than the reference desiva k.. There-fere, for an active fuel length of 150 inebes, a Beral panel length of 144 taches (3 1'aek outback top and betten) satisfies '
the design basis eriticality requirement. ' 4.5.4 Matar Gan Raaeins hatwaan Madulaa 1 The spacing betwoon storage rask modules is moniaally 2.0 inches. Calculattens with CASNO-23 show that apprestaately i 1.32 insbes of water is equivalent ta reactivity effect to the - heral panel between storage es11s. Thus, for normal storage senditions with a 2.0 inch water gap or for the staissa gap of 1.5 inches, the array k.se would be less than the reference desiga k. and heral panels along the walls of the sedules fmetag the water gap would met be necessary. Newever, as an additiemal , and presautionary measure, the rask design provides ier Soral panels on one of the two sedule walls along the water gay. With this senservative configuration, the desiva assures that, even under thaermal er aseident 6. '*sas, the storage reek reno-tivity will remata less than the 0 3i ?,aa regulatory limit. 4 h t 4-8 w-pw--w-y-ow,-- ,----,-:qyy-vgwy--e-+ ye- -w-y-wy-y e----~w-wwwme-
I 4.6 ASNOM!AL AND ACO!0ENT COND T:0NS i 4.6.1 Temeerature and Water Densitv Effects The moderator temperature coeff ::ent of reactivity is negative,and a conservative moderator temperature corresponding to the minimum temperature (6S'T) within the operating range was assumed for the reference design. This assures that the true reactivity will always be lower than the calculated value regardless of temperature or water density. Temperature effects on reactivity have been calculated and the results are shown in Table 4.4. Introducing voids in the water in the storage cells (to simulate boiling) decreased reactivity, as shown in the table. Boiling at the submerged depth of the racks wculd occur at approximately 250'T. 4.6.2 Abnormal Locatien of a Tuel Asrembiv l It is theoretically possible to suspend a fuel assembly l of the highest all:wable reactivity outside and ad acent to the fuel rack, although such a condition is highly unlikely. The exterior walls of the rack modules facing the outside (where such a condition might be postulated to exist) have a stand-off angle that prevents an extraneous assembly item appr: aching 01:ser than 1.4 inches. At this separation distance, :alculati:ns confirm that the reactivity effect of the extraneous assembly is less than 0.0001 ok. Neutron leakage, inherent al:ng the module edge, would signift:antly reduce the reactivity and thereby eliminate the already negligible effect of the extraneous assembly. Thus, the abn:rmal 1 :ati:n Of a fuel assembly will have a negligible reactivity effect, and, if the leakage were included, the rea:- tivity w:uld be less than the reference design basis k.. 4 - 9
-- - ~- - - - - - - . - . - - _ . - - - - . .- - - - . - - ..- -.- -.--.- _
l J l 4.4.3 Benantrie Fuel Ammambly Positionine The fuel assently is aeraally leested La the center of i the storage rask eell with bettes fittings and spacers that mechaatsally restrict lateral movement of the fuel assentlies. i Bevertheless, calculattens with the fuel assently seved into the eerser of the storage rack sell (four-assembly sluster at olesest approach), resulted in a small negative reactivity effect. Thus, 7 the assiaal case, with the fuel assembly positioned ta the seater of the storage cell, yields the mestaus reactivity. I 4.6.4 Drammed Fuel Assembly For a drop en top of the reek, the fuel assembly will some to rest hertsestally on top of the reek with a stataua separation distanee from the fuel of more than the 12 taches ! suffistent to preolude neutres ocupling (i.e., an effectively j infinite separaties). Nasiana espected defernatica under seismie - er assident senditions will met reduce the minissa spesing j between fuel assemblies to less than 12 taebes. ceasequently, fuel assembly drop assidents will met result ta a signittomat increase in reactivity (e0.0001 ok) due to the separatica distanee. 4.6.5 Puel Raek Lateral itevenant Bernally, the individual rask modules in the spent fuel ; pool are separated by a water gap of semina117 2 Laebes in thickness (1 % inch minimum). Lateral metica of 1 fuel reek, postulated as a senseguence of the design basis earthquake, would
- eause saly a steer effeet (e0.4 ineh) en the water gap spasing.
Although this aimer postulated sevement would have negligible reactivity senseguesses, Seral panels are provided alent the , interface between modules which proeludes any reactivity effect regardless of movement. 4 - 10 l l L
I 4 t i f h
4.7 REFERENCES
FOR SECTION 4
- 1. A. Ahlin, M. Edenius, M. Naggblos, "CASMO - A Fuel !
Assembly Burnup Prograa," AE- RF- 7 6 -415 8, Studsvik i report (proprietary). A. Ahlin and M. Idenius, "CA8MO - A Fast Transport Theory Depletion Code f or LWR Analysis," AN8 Transactions, Vol. 26, p. 604, 1977. M. Edenius et al. , 'CASMO Benchmark Report," Studsvik/RF-78-6293, Aktiebolaget Atomenergi, March 1978.
- 2. Green, Lucious, Petrie, Ford, White, Wright, "PSR - ,
63/AMPI-1 (code package), AMPI Modular Code System for
- Generating Coupled Multigroup Neutron-Gamma Libraries from RNDF/S,' OREL-TM-3706, Oak Ridge National '
Laboratory, March 1976. L.M. Petrie and 5.F. Cross, *EENO-IV, An Zaproved Monte Carlo Criticality Prograa," ORNL-4938, Oak Ridge National Laboratory, November 1975. l
- 3. R.M. Westfall et al., "8CALE A Modular Code System for Performing Standardised Computer Analyses for i Licensing Evaluation,* NURSG/CR-020: 1979.
1 1 4. W.R. Cadwell, PD9-07 Referince Manual, WAPD-?M-678, I Bettis Atomic Power Laboratory, January 1967.
- 5. M.G. Natrella, Esperimental Statistics National Bureau of standards, Mandbook 91, August 1963.
4 - 11
. - . . :' "'T':*::':':::mr ':'"**"- ~ " ~ " ' ' ~ ' ~ " ' " " " ' ~ ~ " * ~ ~ ' ' ' ~~~~' '* ~ ~ ~
i I Table 4.1 i
SUMMARY
OF CRITICALITY SAFETY ANALYSES j Temperature assumed for analysis 68'F Uniform Enrichment for Analysis 3.3 % U-235 Reference k (CASMO) .O.9284 ' Calculational Bias 0.0013 Uncertainties Calculational 10.0018 Boral width 10.0017 [ Lattice spacing 20.0048 i SS thickness 10.0004 Fuel enrichment 10.0038 Fuel density 10.0026 Remetal of Flow Channel negative Eccentric Assembly Position negative Statistical combination (1) 1 0.0071 of uncertainties Total O.9297 i O.0071 Maximum reactivity 0.9368 (1) Square root of sum of squares of all independent tolerance effects. 4 - 12
l 1 i i Table 4.2 REACTIVITT EFFECTS OF ASNORMAL AND ACCIDENT CONDITIONS l Accident / Abnormal Condition Reactivity 3ffect Temperature increase . Negative (Table 4.4) 1 Void (belling) Negative (Table 4.4) Assembly dropped on top of rack Negligible (<o.0001 sk) Movement of rack modules Negligible (*0.0001 sk) Misplacement of a fuel assembly Negligible (40.0001.ok) l. 1 , W
* ?
l t' ; e . t P Table 4.3 FUEL ASSEMBLY DESIGN SPECIFICATIONS FUEL ROD DATA i Cladding outsida diameter, in. 0.493 Cladding inside diameter, in. 0.425 cladding material tr-2 ! t UOn. density (stack), g/cc 00 10.294 Tolerance , a 0.216 Pellet diameter, inch 0.414 Barichment (Rack design basis, 3.3 ,. ue,1 form t U-235) WATER RCD DATA l Number of Water Rods 2 Zaside diameter, inch 0.425 Outside diameter O.493 Material tr-2 FUEL ASSEMBLY DATA , Fuel rod array 8x4 o Number of fuel rods 62 Fuel rod Pitch, inch 0.640 Fuel channel, asterial tr-2 Inside dimension, inch 5.278 Outside dimension, inch 5.478 l l 4 - 14
- 1. - . 'T ~ L: . . . . . . . :. . - . _ . - - - - - - - .- -
l 4 I 1 l I i i Table 4.4 i EFFECT OF TEMPERATURE AND VOID ON CALCUIJTED ' REACTIVITY OF STORAGE RACE i Case Incremental Reactivity Change, ok { 48'F Reference 122'F -0.004 174*F -0.013 252'F -0.024 252'F + 20% void -0.067 3 t l i i I t 4 - 15
.l .
.l y .
l. l' o u I l _ 0.075" THCX 80RAL l 1 5.00" 1 0.080" 0.075" SS 90X : m o.ner sPAct e,eias e., 2 g,1g 6 0.08" l D rrui rd, 4 -
= == iun ir e.n.m . .as uir-M.Y. -
se On ot s m. . w e, 5.278"10- O Q G PncH = o.ser OO \ 1 O. O O'0 S ta 8085
. ,, OOOOOOhs a.ar e OOOOOOOh, l
l: i 8.358 1 0.08P MEMlATIIct peces i i INT 10 scmz l .
- - FIGURE 4.1 CROSS SECTION OF SPENT FUEL STORAGE CELL
.~ I 4 - 16
."^^"*
n - .:_,a a - . g- - - . . .= +-na * .a ~ r ->msa - a.* - s - t ,'-[
}
is J l l 1 i 1 4 1 l l
.I 1
'l l
l I l t. F l J 9 L
' i F
APPENDIX 4A 4 ., mg l-l l r 4.
* T*,* 'b * ,! *a n * * , eum a g.7 , ,, . , ... ,, ,,,, ' i*' OT*%% 9 pre -. g, e
4 f
y L-L { 1.0 IN*RODtfC? TON AND StDO4ARY The objective of this benchmarking study is to verify both the AMPX (NITAWL)-KENO IV (Refs. 1 and 2) methodology with the 27-group SCALE cross-section library (Ref. 3 and 4) and the CASMO-2E code (refs. 5,6,7, and - 8 ) for use in criticality safety calculations of high density spent fuel storage racks. Both calculational methods are based upon transport theory and have been benchmarked- against critical experiments that simulate typical spent fuel storage rack designs as realisti-cally as possible. Results of these benchmark calculations with both methodologies are consistent with corresponding calculations reported in the literature. . Results of these benchmark calculations show that the 27-group (SCALE) AMPK-RENO calculations consistently under-predict the critical eigenvalue by 0.0106 2 0.0048 4k (with a 95% probability at 'a 95% confidence level) for critical experiments (Ref. 9) selected to be representative of realistic spent fuel storage rack configurations and poison worths. Siallar calculations by Westinghouse (Ref. 11) - suggest a bias of 0.0120 2 0.0023, and the results of OREL analyses of 54 relatively aclean* critical experiments (Ref. 12) show a bias of 0.0100 2 0.0013. Similar calculations with CASNO-2E for clean critical experiments resulted in a bias of 0.0013 with an ancertainty of 2 0.0018 (95%/95%). CASMD-2E and AMPI-KENO intercomparison calculations of infinite arrays of poisoned cell configurations show very good agreement and suggest that a bias of 0.0013
- O.0018 is the reasonably , espected bias and uncertainty for CASMD-2E calculations.'
A-2 4
. =
,7
! T
{ a, j . -. " j t The benchmark calculations . reported .here indicate that either the 27-group (SCALE) AMPI-KZNO or CASMO-2E calcula-tions are acceptable for criticality analysis of 'high-density j
' spent fuel storage racks. Reference calculations for the rack, i a
designs should be performed with both code packages to provide independent verification. ' u s - L
-i h ,
1 <, L
- u. . .
l 3 i e s1 k i' :f O A-3 lf
. . . ~ . . . . . . . . . - . . , - . .. . .
2.0 AMPY (NITAWL)-KYNO TV aMicWMARK Fit.FULATIONS Analysis of a series of Babcock-& Wilcox~ critical l experiments (Ref. 9), which include scoe with absorbe~r panels typical- of a poisoned spent fuel rack, is . sunsarized in Table 'l I 1, as calculated with AMPX-KENO using the 27-group SCALE cross-
- section library and the Nordheim resonance integral treatment j in NITAWL. The mean for these calculations is 0.9894 2 0.0019, -l c
I conservatively assuming the larger standard deviation calcul- ) ated from the k gg values. With a one-sided tolerance factor corresponding to 954 probability at a 954 confidence level (Ref. 10), the calculational bias is + 0.0106 with an uncer- s tainty of 2 0.0048. Similar calculational deviations reported by Westin-ghouse (Ref. 11) are also shown in' Table 1 and suggest a bias ' of 0.0120
- o.0023 (954/954). In addition, ORNL (Ref. 12) has analysed some 54 critical experiaants using the same methodo-logy, ob*=4aiag a mean bias of.0.0100
- 0.0013 (954/954).
These: published results are in good agreement with the results obtained. in the present analysis and land further credence .to l the validity of the 27-group ANPE-KINO calculational model for use in cri*ia=14ty analysis of high density spent fuel storage L racks. Variance analysis of the data in Table 1 suggeststhe possibility that an unknown factor may be' causing a slightly larger variance than might be expected free Monte Carlo l statistics alone. Bowever, such a factor, if one truly exists,
' is too small to be resolved on the basis of the critical experiment data presently available. No trends in kogg with intra-assembly water gap, with absorber panel reactivity worth, or with. poison concentration were identified.*
*8ignificantly large trends in k.gg with water qmp and with absorber panel reactivity worth have been reportedW) for ANPE-EINO calculations with the 123-group GAN-TEIRNOS library.
A-4 t
I p 1: , U i 3. 3.1 GENERAL The CASMO-2E code is a multigroup transport theory code utilizing transmission probabilities to accomplish two-dimen-sional calculations of reactivity and depletion for BWR and PWR fuel assemblies. As such, CASMO-2E is well-suited to the criticality analysis of spent fuel storage racks, since general practice is to treat the racks as an infinite medium of storage cells, neglecuing leakage effects. CASMD-2E is closely analogous to the EPRI-CPM code- (raf.
- 13) and has been extensively benchmarked against hot and cold critical experiments by Studsvik Energiteknik (Refs. 5, 6, 7 and 8). Reporued analyses of 26 critical experiments indicate a mean kegg of 1.0000 2 0.0037 (la). Yankee Atomic (Ref. 14) has also reported .results of extensive benchmark calculations with CASMD-2E. Their analyses of'~54 Strawbridge and Barry critical experiments (Ref. 15) using the reported bucklings indicate a mean of 0.9987 t 0.0009 (la), or a bias of 0.0013 2 0.0018 (with 954 probability at a 95% confidence -level) .
Calculations were repeated for seven of the Strawbridge and-Barry experiments selected at random, yielding a asan kegg of 0.9987
- Q.0021 (la), thereby confi=ia; that the cross-section library ' and analytical methodology being used for the present calculations are the same as those used in the Yankee analyses.
Thus, the-expected bias for CASMD-2E in the analysis of " clean" critical experiments is 0.0013 2 0.0018 (95%/954). A-5 i e
l-3.2 BENCEMARK CALCULATIONS CASMO-2E benchmark calculations have also been made for the B&W series of critical experiments with absorber panels slantiating high density spent fuel storage racks. However, CASHO-2E, as an assembly code, cannot directly represent an entire core configuration *'without introducing uncertainty due to reflector constants and the. appropriateness of their spectral weighting. For this reason, the poisoned cell configurations of the central assembly, as calculated by CASMO-2E, were benchmarked against corresponding. calculations with the 27-group (SCALE) AMPX-RENO IV code package. Results of this comparison are shown in Table 2. Since the differences are well within the normal EENO statistical vari.ation, these calculations confirm the validity of CASMD-2E calculations for the typical high density poisoned spent fuel rack configura-tions. The differences shown in Table 2 are also consistent with a bias of 0.0013 2 0.0018, datamiaM in Section 3.1 as the expected bias and uncert.ainty of CA8MD-2E calculations. 4
- Yankee has attsmyted such calculations (10) using C&sMD-2E generated constants in a two-dimensional, four-group PDQ model, ob+=iaiag a mean kegg of 1.005 for 11 poisoned cases and-1.009 for 5 unpoisoned cases. Thus, Yankee benchmark calculations suggest that CASMD-2E tends to slightly overpredict roeotivity.
A-6
.ym p- e e t * *
~ ~ . . . .
l J l I 1 REFERENCEE TO APPENDIY A
)
- 1. . Green, Lucious', Petrie, Ford, White, and Wright, "PSR /AMPX-1 package? AMPX Modular Code Systen For y Generating (code Coupled MultLgroup Neutron-GAsuna Libraries from ENDF/B", ORNL-TM-3706, November'1975. Oak Ridge National Laboratory, -t
- 2. L.M. Petrie and N.F. Cross, " Keno-IV. An Improved Monte.
Carlo Criticality Program", ORNL-4938, Oak Ridge ; National Laboratory, November 1975.
- 3. R.M.
Performing-Westfall et. al., "8CALE: A Modular System for Evaluation", 8tanclardized Computer Analysis for Licensing NUREG/CR-02OO, 1979.
- 4. W. E.. Ford,III, et al.,
"A 218-Neutron Group Master! Cross- '
section Library 2or criticality safety 3tudies", ORNL/TM-4, 1976
)
- 5. A. Ahlin, M. Edenius, and E. Eaggblom, "CASMO - A Fuel
- Assembly Burnup Program", AE-RF-76-4158, Studsvik report.
6. A. Ahlin and M. .Edenius, "CASMO - A L Fast Transport Theory a 26,Depletion Code for LWR Analysis", ANS Transactions, Vol.
- p. 604, 1977.
- 7. M. Idenius et ' al. , "CASMO Benchmark Report", Studsvik/RF-78/6293, Aktiebolaget Atomenergi, March 1978 l i
- 8. "CASMD-2E Nuclear Fuel Assembly. Analysis, Ap Users Manual", Rev. A, Control Data Corporation,plications 1982 >
- 9. M.N. Baldwin .et al.,
" Critical Experiments Supporting Close Proximity Water Storage. of Power Reeotor Fuel",- BAW-1484-7, The Babcock & Wilooz Co., July 1979..
- 10. M.G. Natrella, Eseparimental statistles. National Bureau of Standards, Bandbook 91, August 1963.
- 11. B.F. CoEhey #t al., " Comparisons of Ex '
and Calculations IN LWR 8torage Geometries",periments Westinghouse ' \: NFS, las- Tranaaet baa, Vol. 39, p. 531, November 1981. A-7
a ( s
*EFEF.ENCES TO APPENSIX A ICentinuedi i
- 12. R . W .. Westfall and- J. H. Knight, " SCALE System Cross-section Validation with Shipping-cask Critical Experi-I ments", ANS Transactiens, Vol. 33, p. 368, November 1979 13.
"The EPRI-CPM Data Library", ASMP Cermuter Cede Manuals,-
Part stitute, November 1975 CCM3, Electric Power Research In-II,. Chapter 4,
- 14. E.E. Pilat, " Methods for the Analysis of Boiling Water Reactors. (Lattice Physics), YAEC-1232, Yankee . Atomic ,
Electric Co., December 1980.
- 15. L.E. Strawbridge and R.F. Barry, " Criticality Calculatiens for Uniform, Water-moderated Lattices", Nuclear Science and Eneineerine, ~ Vol. 23, p. 58. September 1965.
- 16. S.E. Turner. . and M.K. Gurley, " Evaluation of AMPX-EENO Benchmark Calculations for High Density Spent Fuel i Storage Racks", Nue' lear Science and Encineerine, 80(2):230-237, February 1982. !
l l' t-i 4 A-8
Table 1 RESULTS ' 0F 27-GROUP (SCALE) ' AMPX-RINO IV CALCUIATIONS OF B&W CRITICAL EXPERIMENTS
- at ca ated a We tin ouse r
I 0.9889 t 0.0049 -0.008' II 1.0040 t 0.0037 -0.012 III 0.9985 2 0.0046 -0.008 1 IX 0.9924 2 0.0046
-0.016
-X 0.9907 -o.008 2 0.0039 XI 0.9989 2 0.0044 -0.002 XII 0.9932 2 0.0046 -0.013 XIII 0.pg90 2 0.0054 -o.007 XIV 0.9830 2 0.0038 -0.013-XV 0.9852 2'O.0044 -0.016 XVI 0.9875- 2 0.0042 -o.015 XVII 0.9811 2 0.0041 -0.015 XVIII 0.9784 2 0.0050 -0.015 XIX 0.9988 2 0.0033 -0.016 IX 0.9922 2 0.0048 -0.011 d
XXI 0.9783 2 0.0039 -o.017 Mean 0.9894 2 0.0011(1) 0.0120 2 0.0010 Bias 0.0106 2 0.0019(2) 0.0120 2 0.0010 Bias (95%/95%) 0.0106 2 0.0048 0.0120 2 0.0023 (1) (2) Calculated from individual standard deviations. e=1=_ lated from kegg values and used as reference. A-9 l
t , r . l
; 'i-
[: L . ( 1 Table 2 RESULTS OF CASNo-22 BENCEMARK (INTERCOMPARISON) CALCUIATIONS p A(1) ! B&W Experiment (1) AMPI-KINO IV(2) CASNo-2E 6k
%IX 1.1203
- o.0032 1.1193 0.0010 XVII 1.1149 1 0.0039 1.1129 0 0020 XV 1.1059 2 0.0038- 1.1052 0.0007 Interpolated (3) 1.1024 2 0.0042 1.1011 0.0013 XIV 1.0983 1 0.0041 1.0979 0.0004 IIII 1.0992 1 0.0034 1.0979 0.0013 Mean i O.0038 0.0011 Uncertainty i O.0006 Typical BWR fuel 0.9212 2 0.0027 0.9218 - 0.0006 L rack L
(1) Infinite array of central assemblien- of the 9-assembly v&W critical configuration (Bef. 9). (2) b from AMPX-EERO corrected for bias of 0.0106 Sk. (3) Interpolated- frasi Fig. 28 of Ref. 9 for soluble baron i concentration at the critical condition. K A - 10
{
- 5. 0 ' THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction A primary objective in the design of the high density spent fuel storage racks for the J.A. FitzPatrick spent fuel pool is to ensure adequate cooling of the fuel assembly cladding. In the following section, a brief synopsis of the design basis, the method of analysis, and the numerical results is provided.
Similar methods of thermal-hydraulic analysis have been used in previous-licensing efforts on high density spent fuel racks for ; Fermi 2 (Docket 50-341), Quad Cities 1 and 2'(Dockets 50-254 and 50-265), Rancho Seco (Docket 50-312), Grand Gulf Unit 1 (Docket l 50-416), Oyster Creek (Docket 50-219), Virgil C. Summer (Docket. 50-395), Diablo Canyon 1 and 2 (Docket Nos. 50-275 and 50-323), Byron Units 1 ond 2. (Docket 50-454, 455), St. Lucie Unit one-(Docket 50-335), Millstone Point I (50-245), Vogtle Unit 2 (50-425), Kuosheng Units 1 & 2 (Taiwan Power Company), and Ulchin Unit . 2 (Korea Electric Power Company). The analyses to be carried 'out for the thermal-hydraulic qualification of the rack array may - be broken down into the following categories: i (i) Pool decay heat evaluation and pool bulk temperature variation with time. (11) Determination of the maximum pool local temperature at the instant when the bulk temperature reaches its maximum value. p i i- ' 5-1 l s . . . - . . . . _ _. _ ... . _ _ _ . . . . . , . . . . _ . . . . . . . . . . , . , , .
l p (iii) Evaluation of the =aw % = fuel cladding temperaturoL to establish that bulk nucleate boiling at any location resulting in two phase conditions environment around the fuel is not possible. (iv) Evaluation- of the time-to-boil if all heat rejection paths from the cooler are lost. (v) Compute the offact of a blocked . fuel cell : opening - on the local water and maximum cladding temperature. The following sections present a synopsis of the methods employed to perform such analyses and final results. 5.2 Syntam Description The Puel Pool Cooling and Cleanup System cools and purifies the spent fuel storage pool by passing the pool water - through two heat exchangers, thereby transferring heat to the Reactor < Building Closed Loop Cooling Water System. Water purity
.and clarity in the spent fuel storage pool,; reactor head cavity, and reactor internals storage pit are maintained by filtering and desnineralising the pool water.
The system includes two skimmer surge tanks, piped in parallel, two 100 percent capacity pumps, two 50 percent capacity
. heat exchangers, and two 100 percent capacity filter-domineralizers. piping-and valving have been added to the system.
so that a third heat exchanger can be added if required. Both-puasps take suction from the spent fuel pool shinear surge tanks' common suction header, and pump water threagh two parallel heat
. exchangers to one of two fuel pool filter domineralisers. The 5-2 I
i
, , . . . , . - .-- . - . . - - . - . - . . . -.. . .- ~ . - - . . _ . . -
' ', i
- l..
I I { L l filtered water is then routed to the two fuel pool diffusers, g located at the bottom of the pool. The cooled water traverses.:the pool picking up heat and impurities before starting a new cycle by discharging over the adjustable weirs ' into the skimmer surge tanks.- .
^
\ During refueling operations, the- filtering system is
- operated to maintain the required pool water. clarity in the spent j
fuel storage pool, reactor head cavity, and reactor internals t storage pit. These units may be supplemented with the Reactor
- Water. Cleencp system filter domineralizers (when the pool to l
reactor cavity gate is open), thereby reducing the load on the L fuel pool filter domineralizers. System flow- indication is provided in the common-L discharge header of the fuel pool filter-demineralizers returning to the spent fual storage - pool. A 3-point temperature recorder , gives local system temperatures, thus indicating the performance of the heat exchangers and determining whether the system load E requires operation of supplementary cooling. Differential. pressure indication and alarm is provided for the common filter-demineraliser inlet-outlet. l, p The pumps are controlled from a local panel in the Reactor Building. Pump low suction pressure automatically turns off the pumps. l 5-3 l
. . . . . - . . . - . . . . . . .. - - .. - - -- - -- - ~~ -'
. ~ .
x- , t ,. t Stainless steel piping and valves are installed from the
- filter-deminerali=er discharge to the spent fuel storage pool in order to minimize corrosion product addition to the pool.
3 A spent fuel storage pool level monitor is provided to : alarm abnormally high or low water levels in the pool. Alarms are provided in the Control Room and locally at the spent fuel storage pool' pump panel. , i Level switches on the skimmer surge tanks indicate high, low and low-low tank-levels. High level signal alarms occur only to indicate possible excess water input from other areas. The low
, i.evel alarm informs the operator to manually initiate the makeup water supply. The low-low level signal alarms and stops the spent fuel storage pool pumps.
Makeup water for the system is manually transferred from ; the eendensate storage tanks to the skimmer surge tanks to make up any pool losses. Cooling water for the spent -fuel storage pool heat exchangers is provided by- the Reactor Building closed Loop Cooling Water System. Capability exists to add lake water to the pool through the RER System in the unlikely event of loss of normal makeup system and when pool water level is threatened due o to heavy pool water inventory loss. 1 L All equipment in the system is Q.A. Category II, with 4 1 the exception of the spent fuel storage pool and the level switches which are Q.A. Category I. L 1: I
; 5-4
The characteristics of the two heat exchangers are presented in Table 5.2.1. Alignment of the RER System with the Fuel Pool Cooling and Cleaning System is- available in the event that a full core load is discharged. The combined RER and spent fuel pool cooling system is-used for the full' core offload. The interconnection is sized to handle 1200 gpm of flow from the RER system to the spent fuel pool cooling system..The characteristics of the RER heat exchanger are provided in Table 5.2.2. Since RHR usage would make the LPCI System unavailable, interconnecting the RER and Spent Fuel Pool Cooling System is allowable only during plant shutdowns. Plant technical specifications stipulate ~ that the' RER System may be used for spent fuel pool cooling only when the reactor coolant temperature is below 212*F. A normally closed globe valve is located in the interconnecting piping between the Spent Fuel Pool Cooling System and the RHR System. Fuel pool cooling capacity is increased by using the " standby cooling" feature of the RER System. Any one of the four RER pumps can be used together with one of the two RER heat exchangers for this purpose by opening the normally closed globe valve. Table 5'.2.3 provides additional description of the major equipment included in the Fuel Pool Cooling and Cleanup System. 5-5
The two fuel pool filter demineralizers are located in the Radioactive Waste Building above the waste sludge tank to centralize all water treating processes. Stainless steel pipe is used in the Reactor Building to minimize corrosion product pickup. , Carbon steel pipe is provided from the skimmer surge . tanks to the filter-demineralizers. 5.3 DECAY HEAT LOAD CALCULATIONS The decay heat load calculation is performed in accordance with the provisions of "USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July, 1981. For purposes of this licensing application, it is assumed that the pool contains an inventory of 2784 assemblies accumulated through scheduled discharges for 1977 to 2001 (Table 5.3.1). Further, since the decay heat load is monotonic with reactor exposure time, an upper bound of 1594 full power operation days is assumed for all stored fuel. The cumulative decay heat lead is computed for the instance of scheduled normal discharge 616 in the year 2003. As shown in Table 5.3.2, the ratio of this decay heat load due to previously stored fuel to the average assembly operating power is 0.2483. This decay heat load frem "old" discharged fuel is assumed to remain invariant for the duration of the pool temperature evaluations perfermed in the following normal and full core offloads discussed below. 5-6
5.4 MATHEMATICAL IDEALIZATION OF THE SYSTE_M Two conditions of discharge are considered:-
'(i) Normal discharge (Case 1):
-In this condition, 208 assemblies with 38256 hours of exposure in the reactor are discharged to the fuel pool cooled by two coolers operating in parallel (Figure 5.4.1).
The fuel transfer _ begins 96 hours after reactor shutdown and is carried out at the rate of four assemblies per hour. Table 5.4.1 presents the relevant input data.
~
(ii) Full Core Offload (Case 2): It is assumed that in cycle 16, 144 assemblies were discharged to the pool following the standard discharge procedure. Thirty days later, full core. offload occurs wherein the entire core consisting of'560 assemblies is discharged to the pool. The heat exchanger alignment - for this scenario is shown in Figure 5.4.2. This cooling- mode 'is-maintained until the full core is returned to the reactor. The pertinent _ data for analyzing this condition'is presented in Table 5.4.2. 5.5 MATHEMATICAL MODEL AND RESULTS A number of simplifying assumptions were made which render the analysis conservative. These include: o The heat exchangers were assumed to have maximum fouling. Thus, the temperature effectiveness, P, for-the- heat exchanger utilized in the analysis is the lowest postulated value calculated from heat exchanger technical data sheets. 5-7
O No credit was taken for the improvement in the- film coefficients of the heat exchanger as the operating temperature rises due to monotonic reduction in the water kinematic viscosity with temperature rise. Thus, the film coefficient used in the computations are lower bounds. O No aredit was.taken for heat loss by evaporation of the pool water. O No credit was taken for heat loss to pool walls and pool floor slab. The mathematical formulation can be explained with reference to the simplified heat exchanger alignment of Figure 5.4.1. The basic energy conservation relationship for the pool heat exchanger system yields: dt Ct
- Q1-Q2 dr where:
Ct = Thermal capacity of stored water in the pool t =
=
Temperature.of pool water at time, t Q1 Beat generation rate due to stored fuel assemblies in the pool; Q1 is a known function of time, I from the preceding section. Q2 = Beat removed in the fuel pool cooler This equation is solved as an initial value problem by noting that the cooler heat removal rate must-equal the heat generation rate from previously discharged assemblies. Bence: Weool P (Tin - teoci) = PCONS where the parameters are as fallows: PCONS: Heat generation rate from previously stored
. assemblies Weools Coolant thermal flow rate P Temperature effectiveness of the fuel pool cooler.
Tin Coincident pool water temperature (initial value before beginning of discharge) 5-8 l l
y e l l II teools -Coolant inlet temperature The above equation yields: PCONS I Tin =
+ teool i Wcool P. ]
l The value of Tin computed from the above formula is the initial value of the pool water temperature .(at the start of fuel discharge). ! Figures 5.5.1 and 5.5.3 provide the bulk pool temperature profiles for the normal discharge, and full core offload scenarios respectively. The corresponding heat generation rate profiles are given in Figures . 5.5.2 and 5.5.4 respectively. Table 5.5.1 gives the peak water temperature, coincident time, and coincident heat l' generation rates for both cases. The analyses conducted for the normal fuel pool discharge case are conservative for the reasons presented in Section 5.5 of this report. In addition, a further conservatism is employed relative to the maximum temperature of the ultimate heat sink of the spent fuel pool cooling system. i The spent fuel pool employs Lake Ontario as the ultimate heat sink for removal. of decay heat from the spent fuel pool. The temperature assumed for this sink is 82*F. This maximum value is enployed to coincide with recent plant ~ analyses and safety evaluations which established 82*F as the maximum lake water temperature. This maximum lake water temperature has not been attained and based upon historical records at the plant site, it would not be - anticipated ' that this maximum temperature would occur with any regularity in the future. Furthermore, this maximum lake water
. temperature would occur during an unusually hot summer period and the duration is expected to be brief. In order to achieve the maximum bulk pool. temperature illustrated in this report, the fuel discharge would have to occur simultaneously with a period of maximum lake water temperature. In addition, the duration of a bulk pool temperature in excess of 140*F would occur for only 150 L hours (Figure 5.5.1) assuming the unlikely incidence of a maximum lake water temperature coincident with a reactor refueling.
5-9 L 1
- - - - ._____._x.,... ._ For illustration purposes, Figure 5.5.5 provides the temperature profile if the RHR interconnect were to be removed from the alignment after 30 days of operation, and the pool continues to have the full core inventoinf. The peak pool bulk temperature is seen to reach up to 160'? in a very short time..This indicates that the RER assist mode should be maintained until the full core fuel load is returned to the reactor. 5.6 TIME-TO-BOIL calculations were also performed to determine the time elapsed before bulk boiling of the pool water. occurs if all heat exchanger cooling modes were lost gL the instant when the maximum pool bulk temperature is reached. Results for both cases are presented in Table 5.6.1. It is seen that sufficient time to introduce manual cooling measures exists and the available time is consistent with other BWR reactor installations. 5.7 LOCAL POOL WATER TEMPERATURE In this section, a summary of the methodology, calculations and results for local pool water temperature is presented. 5.7.1 Basis In order to determine an upper bound on the maximum fuel cladding temperature, a series of conservative assumptions are made. The most important assumptions are listed below: 5-10
i i
- o The fuel pool will contain spent fuel with varying time- .
after-shutdcwn (rs ). Since the heat emisrion falls off rapidly with increasing r ,e it is conservative to assume that all fuel assemblies are from the latest batch discharged simultaneously in the . shortest possible time and they all have had the maxisma postulated years of' operating time in the reactor. of each fuel assembly is assumed to be equal and The heat emission rate. maximum. r o As shown in the pool layout drawings, the modules occupy an irregular floor s in the pool.. For the hydrothermal analysis, pace a circle circumscribing the-actual rack floor space is drawn (Fig. 5.7.1). It is - further assumed that the cylinder with this circle as its base is packed with fuel assemblies ~ at the nominal layout pitch. O The actual downcomer space around the rack module, group varies. The nominal downconer gap available in the pool a is assumed to be the total gap available around . the idealized cylindrical rack; thus, the ==ri===' resistance . to downward flow is incorpocated into the analysis : (Figs. 5.7.2 and 5.7. 3 ) (i.e. minimum gap between the- , pool wall and rack module, including seismic kinematic effect). l O No' downconer flow is assumed to exist between the rack:. l modules. L o No heat transfer is assumed to occur between pool water and the surroundings (wall, etc.)_ 5.7.2 Modal _Dancription ' In this manner, a conservative idealised model for the rack-l assemblage is obtained. The. water flow is axisymmetric about the vertical axis of the circular rack assemblage, and thus, the flow is two di====ional (axisymmetric three d1==naional) . Fig. '5.7.2 shows a typical aflow chimney" rendering of the thermal hydraulics model. The governing equation to characterize the flow field in 5-11 l
s + lV _ 4,
+
[1 f4 p4 the pool can'now be written. The resulting integral equation can g.l be solved for the lower plenum velocity field (in the radial f direction) ' and axial velocity (in-cell velocity field), by using
- j. the method of collocation. The hydrodynamic loss coefficients ;
which enter into the formulation of the integral equation are alrn taken from well-recognized sources (Ref. 5.7.1) and where m-discrepancies in reported values exist, the conservative values I are . consistently used. Reference 5.7.2 gives the . details of mathematical analysis used in this solution process. After the axial . velocity field is evaluated, it is a straight-forward matter to compute the fuel assembly cladding temperature. The knowledge of the overall flow field enables pinpointing of the storage location with the =4aimn= axial flow (i.e, = art == water outlet temperatures). This is called the most." choked" location. l In order to find an upper bound on the temperature in a typical cell, it is assumed that it is located at the most choked location. Knowing the global plenum velocity field, the revised-axial flow through this choked cell can be calculated by solving l the - Bernoulli's equation for the flew circuit through this cell. F Thus, an absolute upper bound on 'the water exit temperature and:
==w4=== fuel cladding temperature is obtained. In view of these l , aforementioned assumptions, - the temperatures- calculated in this manner overestimate the temperature rise that will actually occur in the pool. Moltec's earlier computer code.TERRP00L*,. based on the theory of Ref. 5.7.2, automates this calculation. 'The l TEERPOOL has been used in qualifying the spent fuel pools for Enrico Fermi Unit 2 (1980), Quad cities I and II 1981),
Oyster Creek (1984), V.C. Summer (1984), Rancho seco (1983), ( Grand Gulf I (1985), Diablo Canyon I sad II' (1986), among others. I 5-12 i
analysis procedure embodied' in THERPOOL has been accepted by the Nuclear Regulatory Cosaission on several dockets. The.. Code TIERPOOL for local temperature analyses includes the calculation of' void generations. The effect of void on the conservation-equation; crud layer in the clad, flux trap temperature due . to gamma heating, and the clad stress calculation when a void exists,. are all incorporated in TERRPOOL. The peaking factors are given in Table 5.7.1. 5.8 CLADDING TEMPERATURE The marinum specific power of a fuel array qa can be given by: qA-= q Fxy (1) where: Fxy q ==average radial peaking factor specific power fuel assembly The data on radial and axial paaking factors may be found in Table
- 5. 7.1. -
The maw 4mm temperature rise of pool water in the most disadvantageously placed fuel assembly is computed for all loading cases. Baving determined the imaxisma local water temperature in the pool, it is now possible to detamine- the ==winn= -fuel clartding temperature. A fuel rod can produce F -t mes i the average heat emission rate over a small length, where F: is the axial rod peaking factor. The axial heat distribution in a rod is generally a maw 4-m in the central region, and tapers off at its two extremities. 5-13 i
..e. - - ~ , . - - - . ~ - - . - .---..-- . -
-p *
,{
a 1 It can be shown that the power distribution corresponding to the ; chopped cosine power emission rate is given by n (a + x) q(x)'.= qa sin
, 1 + 2a where: i la active fuel length as chopped length at both extremities in the power curve x: axial coordinate with origin at the bottom of the active fuel region The value of a is given by '
1: a= ~ 1 - 2s where: _ _ 1/2 1 1 1 2 '. = - - + n Fs n F2 , y, g2 E where F is the axial r44ag factor. The cladding temperature Te is- governed by a third order-l differential equat;.on which has the fora of d3 T d3 T dT . dx3 + al dx2 l
- og - = f (x) dx where al, a2 and f(x) are functions of x, and fuel assembly geometric properties. The solution of this differential equation 5-14 g - wr
, 1 9
I 1 with appropriate boundary conditions provides - the fuel cladding temperature and local water temperature profile. l l- 1 In order to introduce' some additional conservatism in the i 1 analysis, we assume that the fuel cladding has a crud deposit of l
.005 0F -sq.ft.-hr/ Btu crud resistance, which covers the entire surface.
Table 5.7.2 provides the key input data for local temperature analysis. The results of maximum local pool water temperature are. presented in Table 5.7.3. p (; 5.9 RLOCKED CELL ANALYSIS Calculations are also performed assuming that 50% of the top opening in the thermally limiting storage cell is blocked due to a horizontally placed (misplaced) fuel assembly. The corresponding marian = local pool water temperature and local fuel cladding temperature data are also presented in Table 5.7.3. In all- cases, there is .no incidence of localized-nucleate boiling of tho' pool water. l 5.10 References 5.7.1 General Electric Corporation, R&D Data Books,
" Beat Transfer and Fluid Flow", 1974 and updates.
o 5.7.2 singh, K.P. et al. , " Method for Computing the p Maximum Water Temperature in a Fuel Pool Con *=4a4ag Spent Nuclear Fuel *, Beat Transfer
~
Engineering, Vol. .7, No. 1-2, pp. 72-82 (1986). 5-15
~
I I l Table 5.2.1 SPENT FUF.L POOL COOLER CHARACTERISTICS 1 Number of. coolers in parallel Two Pool water flow rate through each cooler, gyms 375 c Coolant Flow Rate through each cooler, gps: 467 > l. Coolant inlet temperature, 'F 100 Cooler temperature effectiveness, p 0.612 f e 4
?
5-16 i
'_____i____________.__
Table 5.2.2 RER BEAT EXCBANGER DATA Coolant flow rate, gpat 9000 Coolant inlet temperature, 'F: 32 Pool water flow rate, gym 7700 RER Beat Transfer Effectiveness, Ps 0.143 5-17 ' l l 1
'A 6 h
t l l l [ o . 4 s+ Table 5.2.3 FUEL POOL COOLING AND CLEANUP SYSTEM mq EQUIPMENT LIST U Bang Type - centrifugal, horizontal Number - 2 ' Capacity - 525 TPa Total head 259 ft , Materials - 316 88 Raat Exchannara Type - abell and tube Number -'2 Materials - c.s. shell - 304 ss tubes Filtar/Deminary_ liven Type *pewdex" pressure precoat Number - 2 ' Flow rate - 475 gpa Max. pressure drop Filter area - 252 ft 235 psi cation / anion ratio - 2/1 5-18
- - , ., J ., . . _ _ _ .
Table 5.3.1 JAF OPERATING DATA onaratina cvela nimehmenad Puel Cycle Shutdown Assemblies Full Power Assemblies A nata Discharmed operation navn h 1 6/1977 132 410 132 2 9/1978 136 845 268 3 5/1980 160 977 428 4 11/1981 188 1102 616 5 6/1983 200 1133 816 6 2/1985 196 1133 1012 7 1/1987 188 1118 1200 8 8/1988 184 1295 1384 9 3/1990 156 1342 1540 10 10/1991 208 1392 1748 11~ 10/3993 204 1457 1952 12 10/1995 208 1458 2160 13 10/1997 208 1569 2368 , 14 10/1599 208 1594 2576 15 10/2001 208 1594 2784 16 10/2003 144 1636 2928 y 10/2003 208 1280 3136 10/2003 208 680 3344 L l 5-19
-. ' T ' '_ _ _ _ _ _1 _ _ " _l__1_T _~i __i _ ____Z _ _ __ _ _ ___ __ __ . _ _ _ _ _ _ _ _
l 1 J 1
):
1 d J Tahis 5.3.2 ; DECAY POWER AND POOL CAPACITY DATA d Operating Power per Assembly P; Stu/hr 14.85E6 Dimensionless decay power, 0.2483 , SFP Capacity, Stu/'F 2.32E6 Minimum assumed reactor cavity capacity, Btu /'F 2.3236 : I l'
)
e k 5-20 . 4 g , ., a , , ,
.% .-e f .e-*
- e *
. . . ~ . . . ,
l l l 1 I t
, d i
l 1 l Table 5.4.1 h DATA FOR NORMAL DISCRAR32 (CASE 1) l l
.i Number of assemblies 208 I Number of' coolers in parallel 2 :
c Exposure Times, hrs. 38256 l Time of fuel transfer after reactor shutdown, hrs. 96 Fuel transfer speed, time, hrs. 52 i 1 Pool water flow rate, lb/hr 375,000 l l (W in Figure 5.4.1) '
\
l l I r 1 s 1 i 5-21
'J '
. f
?
I i Table 5.4.2 DATA FOR FULL CORE OFFLOAD CONDITIONS (CASE 2) r Number of assemblies in the preceding normal discharge 144 Exposure time of the preceding normal > dischargs, hrs. 39264
- Time between the normal discharge and the l full core discharge, hrs. 720 t
[ Time of fuel transfer of the preceding discharge, hrs. 36 l Number of assemblies in the full core 560 Time of fuel transfer of the full core, hrs. 140 Number of heat exchangers 2 Fuel Pool Coolers ' ' + RER r inter-connect - Time fuel transfer begins after shutdown, hrs. 96 Fuel exposure time in the full core (hrs) 208 assemblies: 38256 208 assemblies: 19128 144 assemblies: 720 Flow rate W 1 x10" , lb/hr (Figure 5.4.2) 6.0 Flow rate W 2 x10" , lb/hr (Figure 5.4.2) 32.5 Flow rate W 3 x10" , lb/hr (Figure 5.4.2) 3.75 Flow rate W 4 x10" , lb/hr (Figure 5.4.2) 38.5 5-22
l i 1; 1 \< l l l l l i
.i 1
i 1 i Table 5.5.1 SFP BULK POOL TDtPDIATURE l l ! coincident ' Beat Ma wi== coincident Gener tion Pool Time Qx10" Temp., ,F (hre) Stu/hr. Normal Discharge 145.84 165.0 13.10 (Case 1) Full Core Offload 133.41 230.0 25.79 (Case 2) e 1 4 e S 5-23
_ . _ . ,. . ._. . . _ _ . _ . . _ _ _ _ _ _ _ . . . _ . . _ , . . _ . . . . _ _ _ _ _ . _ ._-4________._ _ _ _ _ _ _ _ .. _ _..
.i i
i l
-1 l
l 1 1
-)
1 1 i Table 5.6.1 1 Casa Mumber Time-to-Boil thh a ma+me ae oput (Ecurs) i 1 11.86 l 2 5.36' t u i t i l l 5-24
. . . . ~ .--, _ - - - . - , . - , , - - , , . - - ----
'l f
Table 5.7.1 Factor Ymina Radial 1.6473 Axial times radial 2.3633 [ Total 2.4307 P 1 I e s h 1 i l l 5-25 . 9
4 d
)
)
i l l 1 l 1 Table 5.7.2 j l DATA FOR I4 CAL TEMPERATURE . I Type of Fuel Assembly BWR Fuel Cladding Outer Diameter, inches 0.493 i Fuel Cladding Inside Diameter, inches 0.425 Storage Cell inside Dimension, inches 6.06 , Active fuel length, inches 150 : 1 . No. of fuel rods / assembly 62 L Operatiyg Power per fuel assembly 14.845 j
- P, x 10 , Btu /hr
- Cell pitch, in#ses - 6.31 '
l Cell height, inches 170 ( 1 !. Plenum radius, feet 30.0 i l Bottom height, inches 11.0 Min, gap between pool wall 2.00 and outer rack periphery, inches l l l l l 5-26 1 I 2
= ~+ . . _--- - . , . - - -_ _ _ _ _ . _ _ . . -_ _ _ _ . _ _ _ _ . _ _ _ _ . _ _ _ _ -
.-_ - ._ _ _- . _ __ _ . _ _ _ _ . _ . . _ _ _ . _ _ _ _ . . _. . . . _ _ . . _ . _ _ _ _ _ _._~ . _ . _ _ _ . . _ . _ .. . _ _ .
i i I Table 5.7.3 l r NAXIMUM LOCAL POOL WATER AND FUEL CIADDING "T.MPERATURE ! N 50% RLOCKAGE Maximum Maximus Mariana Maximum Local Local Local Local Pool Fuel Pool Fuel Water cladding Water Cladding CAAR M M Tamp.,*F Tamp. *F Case 1 210.9 246.5 226.0 257.5 Case 2 192.8 223.3 205.0 232.9 : e h
?
I G 5-27 e= w j
l i. t t i SPENT FUEL W poot W,40 ' c, T " o W, T W,9 ; \ a Wt,9 Figure 5A.1 Pool Bulk Temperature Mode' for Normal Discharge Scenano 5-28
3 4 4 W 1, T 1 u_ V REAC'CR W1 gpg g CAVITY - ,p] POOL
- l C2,T2 b o
\ P 1,R 1 P2,R2
" i U a W4, T2 W3, T l RHR #+iQl COOLER W3'@A
)
I i
, 1 Wt 1,91 Wt2,$2 i
l-Figure 5.4.2 Flow Configuretion During Full Core Discharge I
'y l
5-29 l
k h 150 _ 2 140 - 'N ; N ! I i _ . 1 t 1 t
. 130- -
u - l o - l = g 120 -- w 110 - i 100- T r rrrrr7 rT rT r TTrrrrf7 rT rTTTTv 3 r7 r7 rrr7T1 O 100- 200 NaO 403
. TIME /tTER FEACTOR SHUTIXMil. Hiti i l
FIGURE 5.5.1 BULK POOI. 'lEMPER ATilRE FOR NORMAI. IllSCII ARGE (CASE 1) l
4 O O [d io ET F L ^ I ~ P k M Fn u iCru. - L F M ** g . b .,* 5 c m e- 5
$ C F
Q u r. L >= , D 3.' E I E io rm m e I N
,n., ec 8' E m
r w Q H l W " I &
- a F c M $
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. F-W <
Y Z E [' F r N w e E i.........n.........................,,[8 g 9 R E G CO S' w 8 2 8 3 8H/018 '31YB NOl1VM3N3D 1Y3H 5-31
150 - 4 . 140 -
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an. Y / N w . N N 13- . / \
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- x N N s N N
^ N <
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~
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110-- r r 1 T r7 r i :-{ i T r7 r r T r7 T r7 r, T r, r rl r, 7 r i 1 O $2) 400 f40 TIME MTER FEACRR SHUiOOWII. IIIts FICURE 5.5.3 BtfLK P001. TEMI'ER ATif RE FOR Fill.l. CONF. OFFI.flAll (CASE 2)
esa.s m.m - -asLva*- m A.- , . -,-4 +a sa ...+na. , , . -2s A - - i 1 f I I ! L. ; I Y ~ , n w
,u-oO w FM *j LYm -
I t)w* = e- (p eZ i
~~
5 I 3 L . > ' I --
- i C
~
f* t= i-
. Qr u a i
>0o= 5 -;=
- C
>= b (*f) 8 i
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bN b i <4 !-
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/ Q i o ce :
r o. iC. i w b- 5
's-k'Os " s td l I W L
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l 4 l r A - i. i W
= as
=. >
U 8IIIII4Iil8iiiI4IIIlIeII4Ii64 4444IIIIIl8843IIIIIlIi33888 44 , oO "- l L E I . 8 3 l; BH/019 '31W NOl1V83N3D IVJH L 5-33
l i t i 160 - t
, . i
- 1 150 b
1 ,- w - a - . w ej 140 -
. 8 -
130 - L 120 - -T r1 T r1--s i i i T rrr rT rvT rvrv T rvT q r r 1 r i 1 t O 400 000 llME AFTER TEACTOR SHUTIX) wit, Hiti FIGilRE 5.5.5 POOL WATER TEMPERATifRE RISE WilEN Rift ASSIST IS REMOVEI) i
.c._. _ . . . . , . - - . . . , . . , . _ ,,,...,,,,,s _m..__ __ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _,._ _. ___,,_.
- 4. - -
9 i! i$i i E! 1 l= h Nh Ni . ki s . k -
/
/
!s
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e IDEAllZATION OF RACK ASSE! ABLY 5-35 FIGURE 5.7.1
i i 5 i l" ' i 1 k Ak
/' /
d7
- TOUT
< - c.,
a T IN g ; 91 % w ad H l l g b $
> s g i O HEAT ADDITION f d ,
( f v I i
./
lf . I / V
/ / /gT IN P,
(w 0 E .] / , i /
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l -l .
.. d THERMAL CHIMNEY FLOW MODEL
. 5-36
, , FIGURE 5.7.2
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PLINUM 1 . 9. CONVECTION CURRENTS 5-37 W THE POOL . l FIGURE 547.3 1 m e y- ,, ,s,-1t'
.,,,,,w,,,,-, g-ymw_* ypy-w.y. , mg ,ew., , , , , - -
. - . .. --. - . - . - - - . - . . - . - ~ - . - - . - -
. l l
I l 1 6.0 RACK STRUCTURAL CONSIDERAfIONS
]
The purpose of this section is to demonstrate the structural j adequacy of the James A. FitzPatrick Plant spent fuel rack design under normal and accident loading conditions following the guidelines of the USNRC OT Position Paper (Ref. 6.12) . The method of analysis presented uses a time-history integration method similar to that previously used in the licensing reports on high density spent fuel racks for Fermi 2 (USNRC Docket No. 50-341), ' l Quad Cities 1 and 2 (USNRC Docket Nos. 50-254 and 50-265), Rancho Seco (USNRC Docket No. 50-312), Grand Gulf Unit 1 (USNRC Docket No. 50-416), Oyster Creek (USNRC Docket No. 50-219), V.C. Summer l (USNRC Docket No. 50-395), Diablo Canyon Units 1 and 2 (USNRC Docket Nos. 50-275 and 50-323) , Vogtie Unit 2 (USNRC Docket No. I' 50-425) and Millstone Point Unit 1 (USNRC Docket No. 50-245) . The e results show that the high density spent fuel racks are structurally adequate to resist the postulated stress combinations associated with level A, B, C, and D conditions as defined in References 6-1 and 6-2. 6.1 ANALYSIS OUTLINE (FOR NEW PROPOSED RACK NODULES) The spent fuel storage racks are Seismic Class I equipasnt. They are required to rou.in functional during and after a Safe Shutdown Earthquake (Ref. G-3). As noted previously, these racks , are neither anchored to the pool floor nor attached to the sidevv.ls. The individual rack modules are not interconnected. Furthermore, a particular rack may be completely loaded with fuel. l. 1 6-1 i b
1 l l l I assemblies (which corresponds to greatest rack inertia), or it may l be completely empty. The coefficient of friction, y, between the l supports and pool floor is another indeterminate factor. ) According to Rabinowicz (Ref. 6-4), the results of 199 tests i performed on austenitic stainless steel plates submerged in water
]
show a mean value of g to be 0.503 with a standard deviation of l 0.125. The upper and lower bounds (based on twice the standard ! , deviation) are thus 0.753 and 0.253, respectively. Analyses are j performed for single rack simulations assemblies with values of l the coefficient of friction equal to 0.2 (lower limit) and 0.8 (upper limit), respectively. In order to predict the limiting j conditions of rack module seismic response, the rack module with ! the maximum aspect ratio (length to width ratio), and maximum mass inertia should be evaluated. Therefore, the 4x14 and 11x12 modules j merit seismic simulation for critical conditions of loading. They l aret o Fully loaded rack (all storage locations occupied), p = 0.sr 0.2 (y = coefficient of friction) o Nearly empty rack (6x14 only) The simulations were performed using normal (unconsolidated) fuelt simulations are also performed for a heavier fuel. These modules are labelled Modules 3 and C in section 2. As stated before, the former was selected due to its largest mass inertia, and the latter due to its maximum aspect ratio. 2-D multi-rack analyses
- are also performed to examine the interaction between racks.
The seismic analyses were performed utilizing the time- . history method. Pool slab acceleration data in three orthogonal directions was developed and verified to be statistically independent. 6-2 P-----p .
- g. - -
,m. . , , , _ . _ . , _ _ _ _ _ _ __,___
b? A The objective of the seismic analysis of single racks is to datermine the structural response (stresses, deformation, rigid l body action, etc.) due to simultaneous application of the three ! statistically independent, orthogonal seismic excitations. Thus, l recourse to approximate statistical summation techniques such as l the " square-Root-of-the-Sun-of-the-8quares" method (Ref. 6-5) is avoided. For nonlinear analysis, the only practical method is , I simultaneous application of the seismic loading to a nonlinear model of the structure. Pool slab acceleration data are developed from specifjed response spectra from two earthquakes: SSE and OBE. Seismic time histories are calculated from the plant response spectra at level : 326.8' at 14 damping. Figures 6.1 - 6.12 show the time-histories and comparison of the corresponding velocity spectra and the > design spectra for ssE and OBE conditions. The seismic analysis of a single rack is performed in three steps, namely:
- 1. Development of a nonlinear dynamic model consisting of inertial mass elements, spring, gap, and friction elements. B
- 2. Generation of the equations of action and inertial coupling and solution of the equations using the
" component element time integration scheme" (References 6-6 and 6-7) to determine nodal forces and displacements.
- 3. Computation of the detailed stress field in the rack just above the baseplate and in the support legs using the nodal forces calculated in the previous step. These stresses are checked against the design limits given in Section 6.5.'
r 6-3
> e -
.t * * . 4 4
. _ . - - . - . . - . .---._..w,.- .m --,-,-r, mecy er z. -v-----. .=~----*e -
, ) I l A brief description of the dynamic model follows. { l 3 6.2 FUEL RACK - DYNAMIC MODEL 1 Since the racks are not anchored to the pool slab or attached l to the pool walls or to each other, they can execute a wide { variety of actions. Fcr example, the rack may slide on the pool ; floor (so-called " sliding condition"); one or more legs may . nomentarily lose contact with the liner (" tipping condition"); or the rack may experience a combination of sliding and tipping , conditions. The structural model should permit simulation of these ! kinematic events with inherent built-in conservatisas. Since the modules are designed to preclude the incidence of inter-rack , impact, it is also necessary to include the potential for inter-rack impact phenomena in the analysis to demonstrate that such impacts do not occur. Lift off of the support legs and i subsequent liner impacts must be modelled using appropriate impact l (gap) elements, and coulomb friction between the rack and the pool liner must be simulated by appropriate piecetiise linear springs. : The elasticity of the rack structure, relative to the base, must . l also be included in the model even though the rack may be nearly rigid. These special - attributes of the rack dynamics require a , i strong emphasia on the modeling of the linear and nonlinear springs, dampers, and compression only stop elements. The term non-linear spring is the generic term to denote the mathematical ' element representing the situation where the restoring force exerted by the element is not linearly proportional to the displacement. In the fuel rack simulation the columb friction interface hatween the rack support leg and the liner is a typical L example of a non-linear spring. The model outline in the remainder l of this section, and the model description. in the following L l 6-4 L
. + . . . . - . . . . . . . . . . - . . . . - . . . ~ . . . . ~ -.,a - - - - - '
w a- v, m-a -
~~, -~~~,,- , --, r-+,s- vm-
i i s section, describe the detailed modeling technique to simulate ' these effects, with emphasis placed on the nonlinearity of the rack seismic response. 6.2.1 Outlina of Medal for cannutar cada DYWimact
- a. The fuel rack structure is a folded metal plate assemblage '
welded to a baseplate and supported on four legs. An odd- l shaped module may have more than four legs. The rack structure itself is a very rigid structure. Dynamic analysis of typical multicell racks has shown that the action of the structure is captured almost completely by modelling the rack as a twelve degree-of-freedom structure, where the movement of the rack cross-section at any height is described in terms-of six degrees-of-freedom of the rack base and six degrees of freedom defined at the rack top. The rattling fuel is modelled by five lumped masses located at H,- .75H, .5H,
.255, . and at the rack base, where H is the rack height as measured from the basa. -
- b. The seismic action of a fuel rack is characterised by i randon rattling of fuel assemblies in their individual storage locations. Assuming a certain statistical coherence (i.e. assuming that all fuel elements move in-phase within a '
i rack) in the vibration of the fuel assemblies exaggerates the i computed dynamic loading on the rack structure. This assumption, however, greatly reduces the required degrees-of-freedon needed to model the fuel assemblies which are represented by five lumped masses located at different levels of the rack. The centroid of each fuel assembly mass can be located, relative to the rack structure centroid at that level, so as to simulate a partially loaded rack. .
- c. The local flexibility of the pedestal is modelled so as to account for floor elasticity, and local rack elasticity just above the pedestal,
- d. The rack base support may slide or lift off the pool floor.
- e. The pool floor has a specified time-history of seismic
. accelerations along the three orthogonal directions.
6-5
i l
- f. Fluid coupling between rack and fuel assemblies, and between rack and adjacent racks, is simulated by introducing
! appropriate inertial coupling into the systen kinetic energy. Inclusion of these effects uses the methods of References 6-4 i I and 6-6 for rack / assembly coupling and for rack / rack coupling l l (see section 6.2.3 of this report).
- g. potential impacts between rack and fuel assemblies are accounted for by appropriate " compression only" gap elements between masses involved.
- h. Fluid dam due to viscous effects. between rack and assemblies, ping and between rack and adjacent rack, is '
conservatively neglected; fora drag, however, may be included.
- 1. The supports are modeled as " compression only" elements for !
the vertical direction and as " rigid links" for transferring horisontal stress. The bottom of a support leg is attached to a frictional spring as described in section 6.3. The , cross-section inertial properties of the support legs are computed and used in the final computations to determine support leg stresses.
; j. The effect of sloshing is negligible at the level of the top of the rack and is hence neglected.
- k. The possible incidence of inter-rack impact is simulated by gap elements at the top and bottom of the rack in the two L
horisontal directions. The most conservative case of adjacent rack movement is assumed; each adjacent rack is assumed to move completely out of phase with the rack being - analysed. This maximises the potential for impact.
- 1. Rattling of fuel assemblies inside the storage locations causes the "gapa between the fuel assemblies and the cell l
vall to change from a maximum of twice the nominal gap to a theoretical sero gap. Fluid coupling coefficients are based on the nominal gap. l 6-6 l l l _- e -a e sw ee .. . . e- e e . . .- e e- w < e . o. - se . ,e. e,
i l
- m. The cross coupling effects due to the movement of fluid from one interstitial (inter-rack) space to the adjacent one is !
modelled using potential flow and Kelvin's circulation j theorem. This formulation has been reviewed and approved by ' the Nuclear Regulatory commission, during the post-licensing multi-rack analysis for Diablo Canyon Unit I and II raracking project. The coupling coefficients are based on a consistant modelling of the fluid flow. Whils updating of the fluid flow coefficients, based on the current gap, is permitted in the algoritha, the analyses here are consarvatively carried out using the constant nominal gaps that exist at the start of the event. Figure 6.13 shows a schematic of the model. Twelve degrees of freedom are used to track the motion of the rack structure, i Figures 6.14 and 6.15, respectively, show the inter-rack impact i springs (to track the potential for impact between racks) and fuel assembly / storage cell impact springs at a particular level. l As shown in Figure 6.13, the model for simulating fuel , assembly action incorporates five rattling lumped masses. The five rattling masses are located at the baseplate, at quarter height, at half height, at three quarter height, and at the top of the rack. Two degrees of freedom are used to track the action of each rattling mass in the horisontal plane. The vertical action of each rattling mass is assumed to be the same as the rack base. Figures , 6.16, 6.17 and 6.18 show the modelling scheme for including rack l elasticity and the degrees of freedom associated with rack , elasticity. In each plane of' bending a shear and a bending spring are used to simulate elastic effects in accordance with Reference 6.6. Table 6.3 gives-spring constants for these bending springs as l l well as corresponding constants for extensional and torsional rack L elasticity. I 6-7
l 6.2.2 Model Descrietion ; j The absolute degrees of freedom associated with each of j the mass locations are identified in Figure 6.13 and in Table 6.1. ) The rattling masses (nodes 1*, 2*, 3*, 4*, 5*) are described by ; translational degrees-of-freedos q7-916* , Ui(t) is the pool floor slab displacement seismic time-history, Thus, there are twenty-two degrees of freedom in the system. Not shown in Fig. 6.13 are the gap elements used to model the support legs and the impacts with adjacent racks. s 6.2.3 Fluid couplina l An effect of some significance requiring careful l modeling is the " fluid coupling effect". If one body of mass (m1) vibrates adjacent to another body (mass m2), and both bodies are submerged in a frictionless fluid medium, then Newton's equations of action for the two bodies have the forms a n . ; (at + N11) X1+N12 X2 = applied forces on mass'a1 + 0 (x1 ) 2 m = .
+M21 X1 + (a2 + N22) X2 = applied forces on mass m2 + 0 IX22) ,
a a X, 1 X2 denote absolute accelerations of mass al and a2i respectively. . M11, N12, M21, and N22 are fluid coupling coefficients which depend on the shape of the two bodies, their relative disposition, ! etc. Frits (Ref. 6-9) gives data for Mij for various body shapes l and arrangements. The above equation indicates that the effect of the fluid is to add a certain, amount of mass to the body (M11 to i body 1), and an external force which is proportional to the i i l 6-8
. . _ . _ _ _ - . - . . ., - _ . - . . - . . . , - . , - - - . . ~ . --
H 3-acceleration of the adjacent body (mass m2). Thus, the acceleration of one body affects the force field on another. This {P force is a strong function of the interbody gap, reaching large valuss for very small gaps. This inertial coupling is called fluid-a coupling..It has an important effect in rack dynamics. The lateral motion of a fuel assembly inside the storage location will
, encounter this effect. So will the motion of a rack adjacent to
~ another rack. These effects are included in tha equations of motion. For example, the fluid coupling is between nodes 2 and 2* in Figure 6.13. Furthermore, the rack equations contain coupling terns which model the effect of fluid in the gaps between adjacent J racks. The coupling terms modeling the effects of fluid flowing between adjacent racks are computed assuming that all adjacent $- racks are vibrating 1800 out of phase from -the rack being- = analyzed. Therefore, only one rack is considered surrounded by a hydrodynamic mass computed as if there were a plane of symmetry located in the middle of the gap region. r , Tne rack-to-rack hydrodynamic mass coupling coefficients Mij are inversely proportional to the- annular gap between the two bodies. This gap is a function of time as the two bodies vibrate, so that the hydrodynamic coefficients Nij are functions of time as _ well. .In the previous equations, the notation 0 (x1 2 )e 0 (X2 2 ) represent additional nonlinear fluid restoring forces that arise from the development of the interbody fluid coupling effects. These nonlinear restoring forces are only important as the gaps between bodies become small as they are also proportional to the
]
inverse of the square of the current gap. Proper accounting of the effect of gap size on the hydrodynamic mass Mij and on the ai 6-9 e rs , . . . . , , , , , , , , , . . . _ _ _ _ . . . . _ , _ . . _ _ _ , _
fluid restorin; forces due to film squeezing is permitted at each step in the dynamic simulation. If the . hydrodynamic mass is conservatively based on the nominal gap, and no updating is included, then those additional geometric nonlinear terms are- not - present. Finally, fluid virtual mass is included 6 the vertical ~ direction vibration equations of the rack; virtual inertia .is - also added to the governing equation corresponding to the-rotational degree of freedom, q6(t) and q22(t). 6.2.4 Da-ine In reality, damping of the rack motion' arises from i material hysteresis (antarist damping), re W,1ve intercomponent-motion in structures (structural damping), and fluid viscous effects (fluid damping). In the analysis, a maximum of 1% structural _ damping as imposed on elements of the rack structure during OBE seismic simulations ahd 24 for SSE simulation. Material and fluid damping due to fluid viscosity are conservatively neglected. The dynamic model has - the provision to incorporate f orm drag ef fects; however, no form drag has been used for this analysis. Subsequent to the completion of all dynamic runs which are reported in Tables 6.5 and 6.6, key (governing) cases were rerun with 1% damping for both OBE and SSE simulations per FSAR requirements. The results show a very minor increase in the equipment response, and all required stress- and displacement limits are satisfied. 6.2.5 Imnact
. Any fuel assembly node- (e.g.,. 2*) may impact the-corresponding structural mass node' 2. To simulate this impact, four com gap elements aroundi each rattling fuel assembly pression-only mode are provided (see Figure 6.15). The compressive loads - developed in- these- springs provide the necessary data to evaluate the integrity of the cell wall structure and stored. array dura.ng the seismic event. Figure 6.14 shows the location of the =
6-10 y p 4 g,g g g, A 5 W- " "
- a
_ .. _ _. , _ ,, _ . . - - m ,. ,; ;. ..;
, .~ .y;;., , ;. p ,, . .-- -, : 3.ng, y- m . .p
. - , - -. ,v- --
t 9' where ~ ' K1 = spring rate of the support leg treated as a tension-compression member
= local cpring rate of pool slab K2
= spring rate of ' folded plate call structure above K3 support leg As described in the preceding section, the rack, along with the base, supports, and stored fuel assemblies, is modeled fu the general three-dimensional (3-D) motion simulation by a twenty-two degree of freeden model. to simulate ":s impact and sliding phenomena expected, up to 64 nonlinear gap elements' and 16 nonlinear friction elements are used. Gap and frictwt elements, with their connectivity and purpose, are presented in Table 6.2.
Table 6.3 lists representative values for the B anc c A Wules used in the dynamic simulations. For the 3-D simulation of a single rack, all support elements (described in Table 6.2) are included in the model.* coupling between the two horisontal seismic abtions is provided both by any offset of the fuel assembly group c6mtroid which causes the rotation of the entire rack and/or by the possibility of liftoff Since inter-rack impact does not occur in the subject modules, only 8 gap elements are used around the bottom and top edges of the rack instead of the twenty described in Table 6.2. Since their purpose is only to signal if an impact occurs, the exact number utilized has no bearing on the final reported results.
. 6-13 1
c t v. * * +p.s,* .g. sp, s. . s.,, f . i *
- i .s y-*.;. r *3 +y* e F* *f! ""'*1*- ' * -
._ _ __ m - , . . . . , - ~ . . _
m
..s .
s-i of one or more support legs. The potential exists for-the rack to be supported on one or more support legs during any. instant of a
- l. complex 3-D seismic event. All of these potential events may be L'
simulated during a 3-D notion and have been observed in the analyses. L 6.4 TIME INTEGRATION OF THE EQUATIONS OF MOTION' l 6.4.1 T4=== History Analvain Uniner Multi-r=ci ii e of Frc dom' Rack Model l l Having assembled the structural model, the dynamic. ! l' equations of action correspm. Jing to each degree of freedom are I written by using Lagrange's Formulation. The systen kinetic energy can be constructed including contributions from the solid structures and from the trapped and surrounding fluid. A single rack is- modelled in detail. The system of equations can be represented in matrix notation as: [N] (q") * (Q) + (G) I where the vector (Q) is a function of nodal displacements and -l velocities, and (G) depends on the coupling inertia and the-ground acceleration. Presultiplying the above equations by (M]-1 , renders the resulting equation uncoupled in mass. ' , we have: (ga) = [x)-1 (Q) + [x)-1 (G} f l-Note that'since the mass matrix can be updated'at every-time step because of the time " varying hydrodynamic effects, the inversion-of the equations is carried out at every increment when
~
the updating option is used. The effect of the previously mentioned nonlinear fluid restoring forces is included in the 6-14 l-l
..7...,,_...,,..,,..,...s..
- .~.........,,,...y.. .c . . . . . . . , . , . . . . . , .,
extent- astheto imply any possibility for overturning.. Run C03' l presents maximum displacements for the case where the horizontal excitation level is increased by 50%. l 6.9 IMPACT ANALYSES 6.9.1 Tmnaet Lo=dina net.==n Fuel Ass ==hiv and call waii ' The local stress in a cell wall is conservatively estimated from the peak impact loads obtained from the dynamic-simulations. Plastic analysis is used to obtain the . limiting < impact load. The limit load.is calculated as 4585 lbs. per cell which is much greater than the loads obtained - from any of the simulations. 6.9.2 Tmnaets netween Adiacent Raeks f
.djacent All of the dynamic analyses assume, conservatively, that-
' racks move completely out of phase. Thus, the highest > potential for inter-rack impact is achieved. The displacements
- obtained from the dynamic analyses are less than 50% of the rack-to-rack spacing or rack-to-wall spacing.
It is also noted that the new fuel racks- do not breach tha theoretical plane between the new racks and the- contiguous existing racks, indicating that impact with existing rack modules-will not . occur. This is a plaus;.ble conclusion in view : of the =i fact that the existing racks.and new racks have markedly different structural characteristics and their displacement time histories will-be randomly phased with respect to each other. J Therefore, we conclude that no impacts between racks or between racks and walls occur during the SSE event. 6.10 MELD STRESSES Critical weld' locations under seismic loading-are at the bottom of the rack at the baseplate connection and at the welds on the support lega3 4 Results from the dynamic analysis using the simulation codes are surveyed and the ==vinna loading is used to qualify the welds an these locations. 6-25 l a . , - . , _-
. . .~ __ _ . - . _ _ - . - . . _ . . ._ . . _ _ . _ . . . _ _ . . _ . . _ _ _ _ _ _ _ . . . . - _ __
l I j i j d 6.10.1 Baaanlata to Rack Walds and call-to-call Walds i Section NF permits, for the SSE condition, an allowable ' I4 weld stress r = .42 Sg = 28,600 psi. Based on the worst case of - all runs repo:-ted, the maximum weld stress for the baseplate to
. l rack welds is 15860 psi for SSE conditions. This value- occurs using a fuel weight of 1200 lbs. per cell. For normal fuel loading the weld stress under SSE at this location is reduced to 10785 psi.
I The weld between baseplate and support leg is checked using limit analysis techniques. The structural weld at that l-location is considered safe if the interaction curve satisfies
- l. F/Fy + Mb /My<1 where Fy , My are the-limit load and moment under direct load only and ' direct moment only. F, Mb are the absolute . values of the
, actual peak force and moments applied to the weld section.-This'is- !- a ' mch more conservative relation than the ac.rual interaction curve. For the worst case simulation, . this criterion gives F/Fy+ < L Mb /My = .409 for the supoort leg to baseplate weld. L The critical area that must be considered for fuel. tube to fuel tube welds is the weld between the fuel tubes. This weld ! is discontinuous as we proceed along the tube length. Stresses in the fuel tube - to fuel %e welds - develop along the length of each fuel tube due to fur assembly impact ! with the tube wall. This occurs if fuel assemblies in adjacent ' tubes are moving out of phase with one another so that impact i 6-26 f I-U
) A b- 8 D M*[ N O 'I#m3 _
,; loads in two adjacent tubes are in opposite directions which would
~'d tend to separate the channel from the tube at the weld. The critical load that can be transferred in this weld region for the SSE condition is calculated as 5056 lbs. at every- fuel tube connection to. adjacent tubes. An upper bound to the load required-to be transferred is
/I x 377.4 x 2 = 1067 lbs.
where we have used a maximum impact load of 377.4 lbs. (from Table 6.5), assumed two impact - locations are supported by each weld region, and have increased the load by /2 to account for 3-D effects. 6.10.2 Heatina of an Isolated call Wold stresses due to hsating of an isolated hot call are also computed. The assumption used is that a single cell is heated, over its entire-length, to a temperature above the value associated with all-surrounding cells. No thermal gradient in the vertical' direction is assumed so that the results are conservative. Using the temperatures . associated with this unit, analysis shows that the v?.ld stresses along the entire cell length do not exceed the allowable value for a thermal loading condition. Section 7 reports a value for this thermal stress. 6.11 SEISMIC QUALIFICATION USING MULTIPLE TIME HISTORIES It is recognized that the time histories corresponding-to a given spectrum are non-unique by definition. Therefore, to provide added confidence in the results, two additional sets of synthetic SSE time histories have been generated to investigate the
- sensitivity of rack behavior to different seismic events obtained from the same response spectrum. Figures 6.20 to 6.31 s - ." '
i 1 ? L E show the additional SSE's together with a comparison of the '
. regenerated and the original spectrums. The events are designated as 2nd SSE (H4, E5, H6 time histories) and 3rd SSE (H7, H8, H9 '
time histories) . Tables 6.7 and-6.4 are similar in content to l Tables 6.5 and 6.6 and present the results of these additional . l y analyses using the two new earthquake sets. While the individual results are different, as would be expected, the conclusions' L .f presented in section 6.4 and 6.9 based on the base time histon analysis remain the same. 6.12 MULTI-RACK ANALYSIS summary of Analysis 1 In order to further confirm the structural adequacy of the racks, a line of modules nas been ' subjected - to a single " horizontal plus the vertical earthquakes to assess the l- implications of multi-rack effects. The model used and the methodology have- been previously used in rack licensing efforts at other dockets, most recently Vogtle Unit 2, and have been approved by the -USNRc. In order to examine maximum rack-displacements, we assume that the 4x14 racks are turned 90 degrees to expose the direction kinmaatically more unstable to j the seismic excitation direction. ' f Figure 6.32 shows the 2-D scenario studied for the five , l' rack array. The following degrees-of-freedom are defined in the model shown here: l x1, 25e X9e X13aE17 = horizontai displacement of I i rattling fuel 22e M6e X10e X14eElg = horisental displacement of mass center of rack 6-28 I l
... -,7.---. - --
,7~-- ,- l
, - . ~ . . -
N
^
i
#3, #7, &11, #15,#19 = clockwise rotation of rack module x4, x6< X12, X16,X20 = vertical displacement of rack-plus fuel ;
The gap elements 3, 4, 10, 11, 16, 17, 22, 23, '28, 29-represent impact springs to track rattling fuel-to-fuel-cell. L impact loads as a function of time. Gap elements 1,.2, 5, 6, 9, 12, 15, 18, 21, 24, 27 and 30 are impact springs used to track , potential rack-to-wall or rack-to-rack impacts. Gap elements 7, l 8, 13, 14, 19, 20, 25, 26, 31, 32 are impact springs to track the vertical load in the support feet in each rack. Each spring represents the cumulative stiffness- of two support feet reflecting the two dimensional nature of the model. Finally, friction elements are used at:each support location s to. simulate the potential for sliding. The limiting lond in each friction element is based on - the instantaneous load in the gap element associated with the support. Fluid coupling associated- with the fluid external to the i racks is included in the model. .The.three dimensional nature of , the external fluid coupling is accounted for by : conservatively as="=4as a larger than actual hydrodynamic gay parallel to the: horizontal direction of the ear *h W = when computing the contribution to hydrodynamic mass due to cross coupling of the. ; action. This conserratively limits the fluid -coupling contribution of the flow in fluid gaps parallel to the horizontal excitation direction and is consistent with the USNRC position in this matter.
- 5-29 1
4 ea b, eu 4 s s e e, * - ==
1 4 b I Fluid coupling between fuel and rattling mass is included. Based on the fuel configuration, we can estimate the kinetic energy of the fluid flow in a conservative manner and include the appropriate coupling effect in the analysis. { The kinetic energy and generalized forces of: the structural assemblage shJwn in Figure 6.32 can be determined and the governing equations- developed by applying the Lagrangian techniques. The USNRC qualified computer code DYNARACK used in the single rack 3-D analysis is then used to study the behavior of the assemblage under the postulated seismic loading for .the , plant. i [ Referring to Section 2 of this report, the particular modules studied - are Modules A (next to the Saath Wall), B1, B2, C1, and C2 (next to the North Wall) . This array is chosen because it contains the largest racks and, a rack with the-largest length to width ratio. This array also has a low rack-ta-wall coupling . contribution, which would nawinise the kinematic response of the racks. As noted earlier, we have turned the C racks 'in this model to expose the weakest 'dir: set ten to an - overturning acaent. All of the cells in all of the racks are assumed fully occupied with normal fuel assemblies. This is the critical case based on the single-rack analysis results. p
- l. The coefficient of friction, p, is .5 and is kept constant through the entire event. This is the mean value of the 6-30 is oe. . 6% +gm , se e eme a we ,r w- ****'a =**"P8' '-#"*
- a 7
r coefficient . of friction expected in the pool. The earthquakes applied'are the N-S SSE and the vertical SSE. A similar model.has been employed in a previcus licensing submittal. for Vogtle Unit , 2. The support feet are modelled by gap elements and the bearing pad areas accounted for in the calculation of the pool , floor stiffness.. Table 6.9 summarizes the design basis values used . in - the i simulation run. Table 6.10 summarizes the results of the regular fuel multi-rack analysis and demonstrates that there are no rack-to-rack impacts or rack-to-wall impacts-. The results show that' the support foot loads are consistent with the single rack sinulations. Figures 6.33 to 6.35 show the time history of the gaps between modules 31 and B2, B2 and C1, and C1 end-C2. Figure 6.36 shows - that the nupport foot movement for rnck B2 is quite small. This is typical for all racks in the analysis. The. kinematic results obtained from the 2-D aulti-rack have the'same orders of magnitude as the 3-D single rack analyses. 6.'13 DEFINITION OF TERNS USED IN SECTION 6.0 81,'82, 83, 84 support designations pi Absolute degree-of-freedon number i-gi Relative degree-of-freedon number i y Coefficient of friction ( 4-31
=_w e -
__w .-m = - - - - - - 1
4 Ui Pool floor slab displacement time history in the i-th direction x,y coordinates- horizontal direction
- z. coordinate vertical direction K
I Impoet spring between fuel assemblies and call K Linear component of friction spring f . Es Axial spring at ~ support leg. locations N compression load in a support foot K R Rotational spring provided by ' the pool. slab Subscript i when used with U or X indicates direction (i = 1 x-direction, i = 2 y-direction, i = 3 s-direction) 6.14 REFERENCES 6.1 USNRC Standard Review Plan, NUREG-0800 (1981). 6.2 ASME P2iler & Pressure Vessel Cede, Section III, Subsection NF (1983). 6.3 USNRC Regulatory Guide 1.29, " Seismic Design Classification,a Mov. 3,,1978. 6.4 "Priction Coefficients- of Water IAabricated Stainless Steels for a Spent Fuel Rack Facility," Prof. Ernest Rabinowics, MIT, a report for Boston Edison Company,
-1976.
6.5 USNRC Regulatory Guide 1,92, " Combining Modal Responses and Spatial components in seismic Response Analysis,a Rev. 1, February, 1976. 6-32
. . . . . . . . ..- . . - . ,.c. . , . . . . . .
.,-,q._.---..,
'1 l
1 6.6 "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering," S. Levy and J.P.D. Wilkinson, McGraw Hill, 1976. 6.7- " Dynamics of Structures," R.W. Clough and J. Penzien, McGraw Hill (1975). I 6.3 " Mechanical Design ~ of Heat Exchangers and Pressure Vessel Components," Chapter 16, K.P. Singh and A.I. Soler, Arcturus Publishers, Inc., 1984. I 6.9 R.J. Fritz, "The Effects of Liquids- on the Dynamic : Motions of Immersed Solids,"-Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167-172. 6.10 " Dynamic coupling in a closely Spaced Two-Body Systea : Vibrating in Liquid Medium: The case of Fuel Racks,"
-K.P. Singh and A.I. Soler, 3rd International Conference on Nuclear Power Safety, Keswick, England, May 1982.
6.11 USNRC Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants," 1973. 6.12 'OT Position for Review and Acceptance of Spent Fuel Storage and Handling. Applications",- dated-. April 14, 1978, and January 18, 1979 amendment thereto. i-I i 1 6-33
. ,. v...~ ..ya,-
..y : -n.- . . . .
.;--_--.- . - - - . . . - ..-. . . .~ - - _ _ .
The results presented in Tables 6.5, 6.6 are representative of the totslity of runs carried out. The ' critical case for. structural integrity calculations is included. Appendix A to this Section 6 ' contains a partial output from one of the DYNAR4CK- - simulation runs of a single rack under D excitation. . The
- initial pages showing input data and model description are given along with the final suasary pages giving maximum loads, I displacements, and stress factors.
2 The results corresponding to SSE give the highest load factors. -However, the results given for the SSE still yield ' maximum stress factors (Ri) below the limiting value for. the CBE '! condition for all sections.The critical load factors reported for. < the support feet are all for the upper segment of the foot ~and for. SSE simulations are to be compared with the limiting value of 2.0. L Results for the lower portion of the support foot are not critical ! and are not. reported in the tables. Analyses show that 'significant margins of safety exist against local deformation of the fuel storage call due to rattling 4 impact of fuel assemblies. t Rest.lts obtained for partially loaded racks will be enveloped by the data presented. Overturning has also been considered for the case of the : C rack adjacent to an open area. This has been done by assuming a. multiplier of 1.5 on the SSE , horizontal earthquakes (more conservative than the OT-~ Position Paper) and ch d imy predicted displacements if there See no obstacles. The horizontal displacements do not grow to such an. r 6-24 _ . . ......~..;-.. .. , ,,. .- ...:~ - - ,> m e .~ ~ . e .~.;. . ~ m.:: . e - - r ,e .m a ~ s v ? :-
V h i extent as to imply any possibility for overturning. the maximam Run C03
- pre =ents displacements for the case where the her!aontal excitation level is increased by 50%.
r b 6.9- IMPACT ANALYSES l - M & Aine Between Fuel Assembly and Call Wall p i 6.9.1 [ The local stress in a cell wall is conservatively estimated from the peak impact loads obtained from the dynamic simulations. Plastic analysis is used to obtain the limiting impact load. The limit load is calculated as 4585 lbs. per cell which is much greater than the loads obtained from any ot ~ the simulations. 6.9.2 Tammets netween Adiacent Raeks adjacent All of the dynamic analyses assume, conservatiwdy, that potential racks move completely out of phase. Thus, the highest for inter-rack impact is achieved. The displacements obtained from the dynamic analyses are less than 50% of the-rack-to-rack spacing or rack-to-wall spacing. ! It is also noted that the new fuel racks. do not breach L the theoretical plane between the new racks and ~ the contiguous existing racks, indicating that impact with existing rack modules will not occur. This is a plaushble conclusion in view of the i fact that the existing racks and new racks have markedly different structural characteristics and their displacement time histories ' j . . will be randomly phased with respect to each other. Therefore, we' conclude that no impacts between racks or between racks and walls xcur during the SSE event. 6.10 WILD STRESSES 6 1 critical vold locations under seismic loading are at the bottom of the rack at the baseplate connection and at the welds on the support legs. Results from the dynamic analysis using the simulation codes are surveyed and the mari=n= loading is used to ' qualify the welds on these locations. 6-25 w;.: v ccs;7v.-. "T: s "' w- ' ' ' ' ~ " '
, . . 7-
L 6.10.1 Bananlata to Rack walda and call-to-call walds Section NF permits, for the SSE condition, an allowable weld stress f = .42 Su = 28,600 pai. Based on the worst case of U all runs reported, the maximum weld stress- for the baseplate to rack welds is 15860 psi for SSE conditions. This value occurs using a fuel weight of 1200 lbs. per cell. For normal- - fuel loading the weld stress under-SSE at this location is reduced to 10785 pai. The weld between baseplate and support leg is checked using limit analysis techniques. The structural weld at that location is considered safs if the interaction curve satisfies F/Fy + Mb /My<1 where Fy , My are the limit load and nosent under direct load only 4.m1 direct moment only. F, Mb are the absolute values of the actual peak force and acaents applied to the weld section. This is a auch more conservative relation than the actual interaction curve. For the worst case simulation, this criterion gives F/Fy+ Mb /My = .409 for the support leg to baseplate weld. The critical area that must be considered for fuel tube to fuel tube welds is the weld between the fuel tubes. This weld
'is discontinuous as we proceed along the tube length.
Stresses in the fuel tube to fuel tube welds develop along the length of each fuel tube due to fuel assembly impact with the tube wall. This occurs 12 fuel assemblies in adjacent tubes are - moving out of phase with one another so that impact 6-26
'-- .. .y~..z. y y .-.--.m..~:.; ; c: :v; r
- - _ -_ _ a - - - _ _ _ - _ _ _ - _ _ _ - _ - _ _ - - - _ _ - _ - - - - - - - _
.rw - - - - - - - - - - - -
z.v T ' ~- ~' ' r
-5 l
l l l loads in two adjacent tubes are in opposite directions which would tend to separate the channel fxmn the tube at the veld. The critical load that can be transferred in this weld region for the ! SSE condition is calculated as 5066 lbs. at every fuel tube connection to adjacent tubes. An upper bound to the load required j to be transferred is l
/2 x 377.4 x 2 = 1067 lbs. l l
where we have used a maximum impact load of 377.4 lbs. (from Table '
' 6.5), assumed two impact locations are supported by each weld region, and ' have increased the load by /2 to account for 3-D effects.
6.10.2 .Mantina of an Imelated call Wald stresses due to heating of an isolated' hot call are also computed. The assumption used is that a single cell is
- heated, over its entire length, to a temperature above-the~value .
associated with all surrounding cells. No thermal gradient in the vertical' direction is assumed so that the results- are conservative. .Using the temperatures associated with this unit, analysis shows that the weld stresses along the entire cell length do not exceed the allowable value for a thermal loading condition. Section 7 reports a value for this thermal stress. 6.11 SEISMIC QUALIFICATION USING MULTIPLE TIME RISTORIES It is cecognized that the time histories corresponding to a given speccrum are non-unique by definition. Therefore, to provide added confidence in the results, two additional sets of synthetic SSE time histories have been generated to investigate the
- sensitivity of rack behavior to different seismic events obtained from the same tesponse sWh. Figures 6.20 to 6.31 6-27 o ,. > ,. . .s .- y . . - ~ .. ...~ . .+ . - - -
,, . z. .
show the additional SSE's together with a comparison of the regenerated and the original spectrums. The events are designated-as 2nd SSE (H4, H5, H6 time histories) and- 3rd SSE (H7, H8, H9 time histories) . Tables 6.7 and ' 6. 8 are biallar in content to Tablas 6.5 and 6. 6 and present the results c.f these additional analyses using this two new earthquake sets. While the individual results are different, as would be expected, the conclusions presented in Section 6.8 and 6.9 based on the base time history analysis remain tha same. 6.12 MULTI-RACK ANALYSIS Summary of Analysis In order to further confirm the structural adequacy of the racks, a line of modules has been subjected to a single horisontal plus the vertical earthquakes to assess the implications of multi-rack effects. The model used and the methodology have been previously used in rack licensing efforts at other dockets, most recently vogtle Unit 2, and have been approved by the USNRC. In order to asemmine maximum rack displacements, we assume that the 6x14 racks are turned to degrees to expose the direction kinematically- more unstable to the seismic excitation direction. Figure 6.32 shows the 2-D scenario srtudied for the five . rack array. The following degrees-of-freedom are defined in the model shown here: x1, x5e Xe 9 X13eX17 = horizontal displacement of rattling fuel x2e X6e X10e X14eElg = horistatal displacement of
. mass cantar of rack 6-28
;l
-.:--,.- <-, e , < ~ c - . .r - . v r. . .- --w -- - - - - ,- ~ ~ ~ ~ + -e ~ - -
"n - ' '
a l i
#,3 97, 911, #15,#19 = clockwise rotation of rack module x4, x a , x12, X16sX20 = vertical displacement of rack 4 plus fuel The gap elements 3, 4, 10, 11, 16, 17, 22, 23, 28, 29 represent impact springs to track rattling fuel-to-fuel-cell-impact loads as a function of time. Gap elements 1, 2, 5, 6, 9,
.l 12, 15, 18, 21, 24, 27 and ' 30 are impact springs used to track
]
potential rack-to-wall or rack-to-rack impacts. Gap elements 7, l 8, 13, 14, 19, 20, 25, 26, 31, 32 are impact springs to track the vertical load in . the support feet in each rack. Each spring l represents the cumulative stiffness of two support fact I reflecting the two dimensional nature of the model. Finally, friction elements are used at each support location to simulate the potential for sliding. The limiting load in each friction element is based on the instantaneous load in the gap ' element associated with the support. Fluid coupling associated with the fluid external to the. i racks is included in the model. Tne three dimensional nature of , the external fluid coupling is accounted for by conservatively; } assuming a larger than actual hydrodynamic gay parallel to.the horizontal direction of the earthquake when computing the contribution to hydrodynemic mass due to cross coupling of the notion. This conservatively . limits the fluid coupling i contribution of the flow in fluid gaps parallel to the horizontal excitation dirertion and is consistent with the Uslatc position in this matter. I 6+29
-_,...y.-.
. ... :. , .. -~ ~ -
e ~
4 1 Fluid coupling between fuel and rattling mass is' included. Based on the fuel configuration, we can estimate the kinetic energy of the fluid flow in a conservative manner and include the appropriate coupling effect in the analysis. The kinetic energy and generalized forces of the structural 1 j assemblage shown in Figure 6.32 can be determined and the-governing equations developed by applying the Lagrangian- ~ techniques. The USNRC qualified computer code DYNARACK used in the single rack 3-D- analysis is then used to study the behavior - of the assemblage under the postulated seismic loading for the plant. Referring to Section 2 of this report, the particular modules studied are Modules A (next to the South Wall), 31, B2, , C1, and C2 (next to the North Wall). This ~ array is chosen because it contains the largest' racks and, a rack with the
~
largest length'to width ratio. This r m y also has a low rack-to-p wall coupling contribution, which would maximize the kinematic ; L response of the racks. As noted earlier, we have turned the C L racks in this model to expose the weakest direction to an overturning acaent. All of the cells in all of the racks are assumed fully occupied with normal fuel assemblies. This is the critical case i based on the single rack analysis results. b L The coefficient of friction, g, is .5 and is kept constant f through the entire event. This is the mean value of ths 1 l 6-30 l
-i coefficient of friction expected in the pool. The earthquakes '
applied are the N-S SSE and the vertical SSE. A similar model has been employed in a previous licensing submittal for Vogtle Unit ' 2. The .l:'toport feet are modelled by gap elements and the bearing pad areas accounted- for in the calculation of the pool-floor stiffness. Table 6.9 summarizes the design basis values used in the [ simulation run. l Table 6.10 summarizes the results of the regular fuel multi-rack analysis and demonstrates that there are no rack-to-rack impacts or rack-to-wall impacts. The results show that the support foot loads are consistent with the single rack simulations. Figures 6.33 to 6.35 show the time history of the gaps between modules B1 and D , 52 and C1, and C1 and C2. Figure 6.36 shows that the support foot movement for rack 52. is quite small. This is typical for all racks in the analysis. The kinematic results obtained from the 2-D multi-rack have the same orders of magnitude as the 3-D single ,ch analyses. 6.13 DEFINITION OF TERMS USED IM SECTION 6.0 S1,'52, 83, S4 Support designations L Pi Absolute degree-of-freedon number i qi Relative degree-of-freedos number i g Coefficient of friction l I i l 6-31 l l
~
, , . . . . - , . . , y ., , . . . , . - . ...
~
-Ui Pool floor slab displacement time nistory in the i-th direction x,y coordinates horizontal direction a coordinate vertical direction K
I Impact spring between fuel assemblies and cell K Linear component of friction spring f Kg Axial spring- at support lee-locations
-c N Compression load in a support "oot K
R Rotational spring provided by the pool slab subscript i when used with U or x indicates direction (i e 1 x-direction, i = 2 y-direction, i = 3 s-direction) 6.14 REFERENCES 6.1 USNRC Standard Review Plan, NUREG-0800 (1981). 6.2 ASME Boiler & Pressure vessel Code, Section III, Subsection NF (1983). 6.3 USWRC. Regulatory. Guide 1.29, e8eismic Design Classification," Rev. 3, 1978. 6.4 " Friction -Coefficients of Water Lubricated Stainless Steels for a Spent Fuel. Back Facility," Prof. Ernest Rabinovics, MIT, a report for Boston Edison Company, 1976. 6.5 USNRC Regulatory Guide 1.92, " Combining Modal Responses and spatial components in seismic Response Analysis,a Rev. 1, February, 1976. 6-32
.- . _ -.. - _ . . . . . . ~ . .. -. - . . - _- . - . . . -. -
l l L 6.6 "The Component Element Method in Dynamics with
- Application to Earthquake and Vehicle Engineering," S.
Levy and J.P.D. Wilkinson, McGraw Hill, 1976. l I 6.7 " Dynamics of Structures," R.W. Clough and J. Penzien, I McGraw Hill (1975). 6.8 " Mechanical Design of Heat Exchangers and Pressure Vessel Components," Chapter .16, ' K. P. Singh and A.I. Soler, Arcturus Publishers, Inc., 1984. l 6.9 R.J. Fritz, "The Effects of Liquids on the Dynamic Motions of Innersed Solids," Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167-172. !. 6.10 " Dynamic coupling in a closely Spaced Two-Body Systen l. Vibrating in Liquid Medium:- The Case of Fuel Racks," - K.P. Singh and A.I. Soler, 3rd International ; Conference on Nuclear Power Safety, Keswick, England, , May 1982. 6.11 USNRC Regulatory Guide 1.61, " Damping values for ! Seismic Design of Nuclear Power Plants," 1973. 6.12 "0T Position for Review and Acceptance of Spent Fuel
' Storage and Bandling Applications", dated April 14, 1978, and January 18, 1979 abandment thereto.
[ l 6-33
=
t 4 l 1 L Table 6.1 DEGREES OF FREEDON Displacement Rotation Location Ur Uy Ug ex Sy og (Node) . 1 P1 P2 P3 44 45 46 \
\
2 p17 pig p19 q20 421 422 Point 2 is assumed attached to rigid rack at the top most point. 2* p7 pg " 3* pg pio 4* Pil P12 5* p13 pt4 1* P15 P16 where: Pi = Si(t) + Ul(t) i = 1,7,9,11,13,15,17 l
= qi(t) + U 2(t) i = 2,s,10,12,14,16,1s
= qi(t) + U 3(t) i = 3,19 Ui(t) are the 3 known earthquake displacements.
6-34
. ..... . ... ... ... . . . . . - - - . . . . . - . -- m -- - --
== ~~ ' " ' ~ " " ~ ~ ~ '
...~. . . . . . . . . . - . _ . . _ _ - - .---. -.- ... ...-. -
Table 6.2 NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS
.I. Nenlinear Enrinen (Gan El===ntal (64 Total) +
Hughgg Mode Imeqtion Descrintion 1 Support s1 2 compression only element 2 Support 52 2 compression only element 3 Support S3 Z compression only element 4 support 84 2 compression only element 5 2,2* X rack / fuel assembly impact
~
element 6 2,2* X rack / fuel assembly impact element l 7 2,2* Y rack / fuel assembly impact element-S 2,2* Y rack / fuel assembly impact element 1*, 3*, 4* and 5* 9-24 other rattling masses for nodes 25 Bottom cross- Inter-rack impact elements
- l. section of rack (around edge) .
Inter-rack impact elements Inter-rack impact elements
. Inter-rack impact elements
. Inter-rack impact elements L .
Inter-rack impact elements-
. Inter-rack impact elements 44 Inter-rack impact elements 45 Top cross-section Inter-rack impact elements
. of rack Inter-rack impact elements
. (around edge) Inter-rack impact elements
. Inter-rack impact elements
. Inter-rack impact elements
. Inter-rack impact elements
. Inter-rack impact elements 64 Inter-rack impact elements l
l 6-35
., _. . , , ;- , ...,- y. - e. 3 - :~; ~>?;,- .
i 1 l Table 6.2 (continued) NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS II. Friction E1333n13 (16 total) EuattAE Nade location Descrintion 1 Support 81 X direction friction i 2-3 support support s1 82 Y direction friction X direction friction 4 5 Support 82 Y direction friction 6 Support 83 X direction friction 7 support 83 Y direction friction Support 84 X direction friction 8 9 Support 54 Y direction friction 81- X Slab acaent 10 51 Y slab soment 11 52 X Slab acaent . 12 82 Y Slab soment l 13. 83 X Slab moment 14 s3 Y Slab soment 15 84 X Slab soment 16 S4 Y Slab noment l l l 6-36
, - 3_ - ~- w. 7- -- 3 ,; ,,: ;.~~; - . , ~ r. :; c - <-- - - '
l l 1 l i l Table 6.3 l
)
TYPICAL INPUT DATA FOR RACK ANALYSES-(lb-inch units) Module B Modula C ! Support Foot Spring 4.37 x 106 4.41 x 106 Constant Ks (#/in.) Frictional Spring 1.061 x 108 1.061 x 108 Constant Kg (#/in.) , l_ -Rack to Fuel Assembly .409 x 106 (x) .382 x 106 (x)
-Impact Spring Constant (f/in.) .487 x 10 (y) 2.13 x 10 {y) l Elastic Shear Spring for 56421. (x) 8097 (x)
- Rack (#/in.) 79507. (y) '218963. (y)
, Elastic Bending Spring 8.244x109-(y) 1.60 x 109 (y) for Rack (#-in/in.) 9.798x109 -(x) 8.469x 109 (y) Elastic Extensional Spring 1.99 x 107 - 1.267 x 10 7-(#/in.) Elastic Torsional Spring 1.83 x 108 1.087 x 108 l ( #-in./in. ) Foundation Rotational Resistance 4.586:x 107 .4.586 x 107 Springs KR (#"iA*/iB+) Gaps (in.) (for hydrodynamic calculations) (hi , h3 are ,x faces; and h1 1.25 2.5 h,h4 2 are ,+ y faces, respectively) h2 .75 .75 i h3 2.5 10. h4 .75 10. 6-37 7 . , , . ,v , , , . 7 , v: 7. w;. ---- . ; e : , ?; ; . - .. : Y ~ _ _ _ _ _ _ _ _ _ _ _ _ _ ' " _ _ _ _ '_;~*_'_______ _ _ _ __'*' _' __
~ . _ _ _ _ _ _ _
. 7-t Table 6.4-RACK MATERIAL DATA Young's Y1 eld U1timate Modulus Strength Strength Material E (psi) Sy (psi) Su (psi)--
304L S.S. 27.9 x 106 23150 68100 l_ Section III Table Table Table Reference I-6.0 I-2.2 I-3.2 i l I SUPPORT MATERIAL DATA Material l l 1 ASTM-240, Type 304L 27.9 x 108 23,150 68,100 l- (upper part of support psi psi psi ' feet) 2 ASTM 564-630 27.9 x 100 101,040 145,000 psi psi psi l 6-38
.v + .
I e h TABLE 6.5 STRESS FACTORS AND RACK TO FUEL IMPACT LOAD l
- l. STRESS FACTORS Rack / Fuel Impact Lead (f)
Run Remarks (Pet Mll) R1 R2 R3 Col Rack C 377.4 .022 .017- .133* Full Load _ 6x14 Heavier Fuel .224 .041 .175** Cof = .8, SSE
- CO2 Rack C 339. .015 .011 .101 l SSE, CCF = .8 Full Load, .159 .027 .168 Regular Fuel C03 Rack C N/A N/A N/A . N/A !
CCF = .8 (Wot applicable ! Full Load . for this case) l t Heavier Fuel ! 1.5 SSE in horizontal -J directions (stability check) C04 Rack C 63. .008 .003 .036 SSE, C0F = .8 8 Cells centrally .037. .007 .064 Loaded, Regule.r Fuel B14 Rack B1 Negligible .019 .006 .004 ' Cof = .8 Impact load
-Full Imad, OBE 11x12 .220 .012 .084 Heavier Fuel Upper values are for rack beseplate section.
Lower values are for support foot cross section (upper part) See continuation of table for stress factors R 4-R7) . 6-39
. . . . . .. - -- .,-...-<3.- -a
.* we *a.-r-** ' v "<* */ "'"; * * #
I
- ~-
.. . . . . - - - - - - - . - ~ . . . .- - . .- . . . -
i l Table 6.5 (continued) i 1
, STRESS FACTORS I Rack / Fuel Impact Load (9) l Run Remarks (Per Cell) R1 R2 R3 B13 Rack B1 245. .026 .012 .109* 1 Cof = .8, SSE ;
Full Load Heavier Fuel .325 .023 .209** B12 Rack B1 124. .016 .008 .050 Cof = .2,.SSE _ Full Load, Regular Fuel .172 .019 .206 Bil Rack B1 Negligible .014 .004 .003 4 Full-load, Impact Load i Regular Fuel .303 .162 .397 Cof = .8, OBE 4 I l. B10 Rack B1 124. .016 .008 .050 Cof = .8 Full Load, SSE .172 .019 .206 ! .6- Regular Fuel
' Upper values are for rack baseplate section.
Lower values are for support foot cross section-(upper part) See continuation of table for stress factors R4-R7 ). l 6-40
.. . ._ g .
,.,...t,,. ,
7 , . -
~. . _ . _ . . _ . - _ . _ _ - . - - . _ . . _ . . _ _ _ _ _ _ _ _ _ .
l Table 6.5 (continued) h STRESS FACTORS L- Rack / Fuel l Impact Run Remarks (Per Cell) R4 R5 R6 R7 Col See previous pages .139 .196 .229. .018' for these columns
.254 .407 .443 .025 L
CO2: .074 .144 .169 .015 l , .169 .262 .293 .026 l l C03 N/A N/A- N/A N/A i t C04 .01 .042 .048 .004
.051 .089 .099 .01 B14 .004 .020 .021 .006
.088 .271 . 282 .012 l
B13 .082 .151 .177 .017
.159 508 -.539 .032 512 .035 .072 .084 .012 I
.129 .334 .365 .030 i
i' l [ 6-41
,. - , . .. . . ; . . , . . ,,p . y . ,,. pg_; -. y;, . ,...-
. ,7.y, .. m . . :. .:-. . .
. _a.-, - - , .--.~
o li l: r I, l-L l 1 Table 6.5 (continued) i STRESS FACTORS Rack / Fuel ,' Impact ' Run Remarks (Per Call) R4 R5 R6 R7 L-Bil See previous .003 .016. .017 .004 pages for these l , , columns '
.061 .162 . 170 .008 B10 .034 .072 .084' .012! i t.
l
.103 .298- .322 .024 l- -
1 .; l I l i e e e h { 6-42
. ,,w. .. y. -.. .a,....,..,..,.,..,.
,,,.7- - -
- r. ~- ~ ~. - --~
- .--w
i Table 6.6 RACK DISPIACEMENTS AND SUPPORT LOADS (all loads are in 1bs.) rLooA Lo&D x&xzzou (sua of all SUPPORT VERR 2 CAL sEEAR DX DY l RUN"" support feet) LO&D LO&D' LO&D" (in.) {in.) ] col 1.27x105 1 44990. 52000 4147 .6682 .0011 " ' l 2 43700. 30502 6787 .0026 .000s .l 3 40940. 4 52000. CO2 7.16x104 1 2.75x104 34610. 1018. .3826 .0626 I 2 2.840x104 11693. 4313. .0017 .0007 3 3.265x104 { 4 3.461x104 l l C03 N/A N/A N/A N/A 1.039 .1071 )
.0043 .0011 l C04 1.621x104 1 0.253x103 8253. 977. .0500 .0258 2 7.496s103 7393. 1454. .0032 .0009 !
3 6.8343103 i t' 4 7.073s103 l'
~ The first line in any set of _ data is the masiana vertical load and the second line reported is the vertical load when the not horisontal shear at the liner is maximum.
The first line is the met horisestal liner shear when the vertical ' load is
==4 ==; the eeooed line is the masiana valas of the not horizontal shear on any single support foot.
The first line reports roeults at the top of the raok; the second line reports results at the baseplate; the times at which these ==4- occur any be different. l "" see table 6.5 for definition of runs. 6-43 1
. , . . . g ,.,, z - -.n. . . r . . . .
.o . . - m ,.-,,.,.y,.y.,.,.;-. -c,.. y--.- . . -, -- .
1 l 1 1 l l Table 6.6 (Continued) l RACK DISPIACEMENTS AND SUPPORT LOADS (all loads-are in-lbs.) , floor LQhD MAXIMUM (sua of all scFroRT UntrICAL SEEAR DE DT .l RUN supper'i: 2eet) LOAD LQ4D* LOAD" (in.) (in.) r 314 1.924x105 1 48130. 48140. 5 5. - .000 . 0072 2 48000. 40441 1803. .0000- .0000 3 40090. 4 48140. 313 2.046x105 1 59820. 71120. 4871. .1943 .2059 2 71120. 34439 4873. .0011 .0014 3 61640. 4 70000. l. l -. 311 1.150x105 1 36820. 37470. 2404. .0798 .0939 l 2 34240. 33715. 4459. .0005 .0006 3 37470. 4 32410. all 1.131x105 1 28300. 20300. 46. .0058 -.005 2 20260. 23053. -1255. .0000 .0000 3 28270. 4 20300. 310 1.150x105 1 34820, 37470. 2223. .0797 .0936 2 34190. 20548. 3599. .0005 .0006 3 37470. 4 32300. 6-44
...,.:,..~.... ,a.-,.e,.~. - . .. :. - . ,, -
- . .. -. . . .. - . - ~ ~ . - . . .- - . . . - .
1 l l TABLE 6.7 STRESS FACTORS AND RACK TO FUEL IMPACT LOAD STRESS FACTORS
" Rack / Fuel Impact Icad Run Remarks (#/ Cell) R2 R3 '
R1 b20 11x12, Full 141 .015 .009 .036 Normal . 2nd SSE .174 .018 .115 I Cof. = .8 l C21 6x14, Full 373 .021. .ul? .147 Heavier Fuel 2nd SSE .199 .018 .199 Cof. = .8 C22 6x14, Full 304 .014 .009 .111 Normal 2nd SSE .146 .022 .150 Cof. = .8 C34 6x14, 4 Calls 55. .008 .003 .032 with fuel (normal) 3rd SSE .045 .007 .079 Cof. = .8 C32, 6x14, Full 226. .014 .009 .127 Normal 3rd SSE .147' .022 .207 Cof. = .8 C31 6x14 381. .021 .015 .151 Full Heavier Fuel .236 .028 .280 Cof. = .4 C24 6x14, 8 Cells 85. .009 .003 .033 with normal fuel .037 .006 .070 2nd SSE ' Cof. = .4 a l l l 6-45 l
+
e w wg e m r.e y e, < <-.e-a- .y ;*- .--o- e y -o +~ - - e,~ *e-e- - .++ , o ~s ~ , - , - - ~ ~ ~ , - - -
i Table 6.7 (continued)- STRESS ZACTORS Rack / Fuel . Impact Load ' Run Remarks (f/ Call) R2 R3 R1 b32' 11x12, Full 159. .015 .012- .066 Normal fuel 3rd SSE .204 .038 .252 Cof. = .2 b30 11x12, Full 159. .015 .012 .066 Normal fuel ; 3rd SSE .205 .028 .387 ! Cof. = .8 b23 11x12, Full 239. .022 .015 .072 Beavier fuel 2nd SSE .309 .030 '199 Cof. = .8 b22 11x12, Full 141. .015 .010 .036 Normal fuel 2nd SSE .174 .022 .115 Cof. = .2 6-46 J
, , e a t' * *' S' *t" *
- L
/
d Table 6.7 (continued) STRESS FACTORS Rack / Fuel Impact
^'
Run Remarks (Per Call) R4 R5 R6 R7 b20 See previous pages for .060 .071 .082 .009 these columns
.120 .254 .272 .017 t.
I L C21 .071 .138 .161 .023 L. .
- .118 .364 .394 .030 C22 .062 .134 .156 .013
.139 .259 .283 .024 C34 .018 .046 .053 .004'
.043 .118 .131 .012 C32- .077 .139 .162' .015
~
.139 .290 .317 .032 l
L C31 .120 .199 .233 .027
.193 .406 .447 .040 -
1 C24 .004 .040 .046 .004 )
.043 .096 .106 .043 1 l
6-47 '
5'e t p Table 6.7 (continued) Rack / Fuel Impact Run Remarks. (Per Cell) R4 R5 R6 R7 , b32 See previous pages .077 .114 .133 .011 for these columns
.257 .444 .490 .037 b30 .073 .111 .130 .012
.139 .356 .387 .025-s b23 .097 .118 .137 .015 i
.202' .443 .468 .029 l
! b22 .060 .071 i
.082 .009
.151' .276 .298- . 017-L t
I i I 6-48
a 1 Table 6.8 RACK DISPIACEMDfTS AND SUPPORT LOADS < Tor Additional Seismic Loads (all loads are in lbs.) ; t FT00R Lo&D MhEIMUM (sum of all SUFF00tT VERTIC&L SEE&R DE DT RUN support feet) LohD LotD' Lo&D*' (in.) (in.) (x 105 ) (x 104 ) i l b20 1.123 1 3.269 30120. 1418. .1192 .0652 2 3.754 20304. 2640. .0000 .0005 , 3 3.333 4 3.812 C21 1.253 1 4.159 44210. 1551. .3466 .0894 2 4.103 16797. 4343. .0013 .0009 3 4.421 . 4 4.250 l C22 .71 1 3.032 31790. 1476. .2057 .0671 i 2 3.020 24354. 4012. .0011 .0007 3 3.179 4 3.072 C34 .1495 1 .9796 9796. 1954. .0904 .0254 l 2. 9121 9796. 1954. .0071 .0044 l 3 .6759 4 .7205 ; C32 .7054 1 3.162 32280 2439. .3859 .0770 2 3.079 20463 4621. .0014 .0008 3 3.220 4 3.197
~
L l 6-49
- - - .. . . - - . . . . . ~. .. .. . . _. - -
Table 6.8 (continued) , L ' RACK DISPLACEMDITS AND SUPPORT LOADS For Additional Seismic Loads-(all loads are in 1bs.) FLoca LonD narzmx (eum of all SUPPORT VERTzcAL SEEAR DX D1f l RUN suppo ieet) Lo&D LQ&D' LQhD" (in.) (in.) l
-(x10p) (x 104 )
c31 1.243 1 4.823 51670. 3484.- .5984 .0918 l' 2 4.881 36229. 5857. .0023 .0010 3 5.000 4 5.167 C24 .1773 1 .741 7993. 1598. .0617 .0040 2 .7467 7993. 1598. .0347 (3108 3 .7993
- 4. .7574 b32 1.116 1 4.462 44620. 4497. .181 .1215 2 4.353 39074. 6658. .0011 .0008 3 4.415 4 4.473 b30 1.116 1 4.455 44730. .2269 .1732 .1215 2 4.353 38396. .4554 .0010 .0000 t
3 4.380 4 4.473 - l b23 1.986' 1 6.112 67470. 2547. .2287 .1309 2 6.590 49234. 4333. .0013 .0009 3 5.869 4 6.747 b22 1. 123 1 3.27 38120. 1418. .1192 .0652 2 3.754 32004. 3234. .0000 .0005 3 3.340 4 3.812 6-50 i 1 l
, -. - - .. . - .. . . . . - . - . . .. .. - ~ .. .-
l 1 l L. Table 6.9 1 l spring constant values for Multi-Rack Analysis Rack-to-Fuel (A,B 6 Gap Elements .409 x 10 f/in racks) . Support Foot D Gap Elements .882x107 #/in. Friction Elements .212 : 1010 coefficient of .5 Friction Rack-to-Wall .1 x 106 (top of rack) Impact Springs .2 x 10 (baseplate to wall) Rack-to-Rack .05 x 1Q6 (top) Impact Springs .1 x 10' (baseplate) Rack Reight 171.* Support Foot Reight 11.625"
& B::1 B:ll1 E-1 E-1 width of each 70.25 70.25 70.25 89.375 89.375 rack (in.)
Length of Rack 70.25 76.625 76.625 38.5' 38.5 (in.) ., (parallel to horizontal ' direction) Rack Weight (lbs) 12800 13900 13900 8850 8850 Fuel Assembly Weight 643 lbs. per cell Side Gape for Fluid 7.5" cross coupling 6-51
. - . . - . . . . . . . . . . . . . . - . - - . . - - - - . . - - . ..n.c.--,. .
~~- ~ < . - - , ~ .s.m-
. , - - ~
I
.I Table 6.10 Results of Multi-Rack Analysis
-f Maximum Vahtas Rack-to-Rack Impact Force No Impacts Rack-to-Wall Impact Force No Impacts Support Foot Loads Rack A 51950. lbs.
L Rack B1 58450. lbs. Rack B2 57400. lbs. I Rack C1 34200. lbs. . Rack C2 34245. lbs. Upper Bound on Displacement at Top of Rack Rack A .0733" Rack.B1 .08129" Rack 52 .08149" Rack C1 .07032" l Rack C2 .06524" l Call-to-ruel Assembly Impact Load Per call l' Rack A 82.4 lbs.
- . Rack B1 94.7 lbs.
i Rack B2 113.3 lbs. Rack C1 82.2 lbs. Rack C2 74.9 lbs. 1 1 l l 6-52 l: l
,. . . . - . .. . . . . - . - . . ~. . . . . . - . - - . . - - . . . . . - . - . . - . - . . . - . . f o i L o o I I I __6 1 l l 1 -o , I I l l 1 -O
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- 1.00 ~ . , , . ,
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-2.50 , , , , , , , , , , , , , , , , , , , , , '. ' 1 10 10
- Freq. (Hz.) ,
FictRE 6.12
. _. . - - .-. . . . . . . . _ _ . - - . - . _ - ~.
r t
@ Fei 'i t e ,
'A 2 ' ' ~7 !
/ '
Ot /2* Ie P8 i e, 97 fl . \' H/4 ;
- e 3*
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'8 s
CINTER LINE N h/4 H ,, ,.
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l P11 m
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LONG DIRECTION k 8
/ '
SUPP'T "' ~
/f l- S } S.g .
44 TYP. FMICTION 4 ELEMENT h SCHEMATIC MODEL FOR OYNRACK 6-65 N 6.13
, . . _ . . _ ~ . . . . . . . . . . . . . , . . - . . . . , . , , , . . . . . . . , . _ , . - . . . . . . . . , _ . . . . . . , . . , . . . , . . . . .
. . _ . _ _ _ _ _ _ _ . _ . _. . . . . _ _ _ _ . _ _ __ _ _ . . _ . _ . . _ . . ...~ . _ .- __ _.,
TYPICAL TCP IMPACT ELEMENT f; s W
,(
,W $4 -
M & - I f g g < {[ RACK STRUCTURE TYP. EDITOM IMPACT
- ELEMENT P ,
s mm A m, . M l /7/7 r RACK TO.AACK lt.tPACT SPRING 3 6-66, ' FIGURE 6.14
-. .. . .- . . .. . - - . - _ . . - _ - . _ - - . _ , . ~. --. -. -.
l
-?
I I
- e t
i !. I I 1 Y i t
/ CELL WALL ,
P ' F : MA!!! XE '. A --
,,is v,7 _
/
FUEL ASSD'ELY! CELL IMPACT $7 RING
,/ .
/
/
Y - E I
/
l /
/ -
l l
/
/ m I. eX .
l l IMPACT SPRING ARRANGEMENT 6-67 AT NCOE I FIGURE 6.15 1
l. i i
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/ /
W< u. l* . j . i
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IR r
/
I l ( Y i
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i FIGURE 6.16 . 1 i DEGREES OF FREEDOM MODELLINC RACK MOTION i ( I I
4: t t
- 2 l
i! > , 1 7
-e
-l q
! 17 I '..
i L 9 , , . 1 le i ' - l l L , , 2 .
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FIGURE 6.17 4 RACK DECREES OF FREEDOM FOR I_Z PLANE BENDINC' i t b l I
+ T- = -.- - - - -- 4 -+ c_,_ -+ ,- +__._u ,.,_,___.2___m_________...__.mm__ _ _ _ _
? 's .
t 1 9 18
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2
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. - f ,
w O 20 9 L ( 2 \ :
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( C ' 8 1 ed } O l i i l
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- FIGURE 6.18 ,
RACK DEGREES OF FREEDOM FOR Y-Z PLANE BENDING O
- . . . . + e .-n.z, .e-- .a e. y,. ,a. .w,., , - . . - - ,,
- s. ..-_:n,,e.. .,,--e
R'EL ASSY/ CELL ; IMFACT'SPRINC . K Mi hh o j l l 0.25H i l l
, \
F/2 s 1@hY 'I I o , 0.23H FaCK N C.C. 1' 5 s -41gp 1' , t u ; 0.25H -
, H/2 s i 4m v .
1 TYPICAL PJ U LING MASS 0.25H s FRICUc:! IhTERTACE -
$CY "
f' g , L S SPRING, I . f
$ r
- Mt 4
,, f b
FOUNDAD0N ROTAD ONAL CQ7LIANCE
, snuC, E, .
s FIGURE 6.19 2-D VIEW OF RACE MODEL 6-71
O o d r5e ~
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_ FITZPATRICK SSE SPECTRUM COMPARISOIl 1-11
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FITZPATRICK SSE SPECTRUM COMPARISON H5
-5.00 , . , , .
1 1 10 ' i freq. s HZ') e i FICilitE (e.21
. .-. . . - . _ . , - . . , , . , _ . . ~ . _ . . . - - . . . . , . _ . . . . . . .
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- SE og FITZPATRICK SSE EARTilOUAKE ACCELERKilON H6 Vertical I 1
g g . . . . . . hdds ' ' ' ' "'~Eds'"~ ~ '~Mddo' ' ' ' ' ' i Moo ' 1 Time Steps (.01 sec./s te ,) .3 FICllRE fi.24
-~ . - . , . -. . _ . . . . _ . . . . _ . - _ _ _ . . . - -
i 9.00 - ,
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FITZPATRICK SSE SPECTRUM COMPARISOf f HC - Vertical -
~ ;
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FITZPATRICK H7 SSE EARTHOUAKE - Horizontal I
- 5.00 , , , , , ,
, , , r- . ,- , ,
r . . ,__ ,,, i 1 1D 2 10 freq. (HZ) . FICilRE ( 27 : i
. - - . _ - ~ _ . - _ . . . - -
O ^i 1 i
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~
o ! o ! i l ! l i' o FITZPATRICK SSE EARTHQUAKE ACCELERATION H8 - Horizontal ; vs ,
- g.......g,j...... ,j..,....g , ,
! Time Steps (.01 sec./ step)
- n - e <. . u
4
~
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t 85 . ' a 05 O Q)d.- c4 i 7 (f) ^ f f' , co,: < f ~
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8 7 %.- \ o- - l . l g- RTZPATRICK SSE SPECTRUM COMPARISON H8 - Horizontal j Q u 5 5 5 5 5 s s l 8 s y a y y s]
"' 1- 10 10
- FREQUENCY (Hz.)
FICllRE 6.29 I______________ -- _ -
. _ _ . _ _ . . ~ ,_ _ . . . , , .. ._ .,. - . . . . ~ . . . . - . . . - . . . . . . . . . . .
1
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o: ' l:~ o FITZPATRICK SSE EARTHQUAKE ACCELERATION H9 - Vertical N. 9.m.........,.........,.........,........., *m.m am.m 12m.m ism.m Time Steps (.01 sec./ step) F10llitE 6.30
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Pool Liner ' i - Dan.it-RACK CDatt., S Rncoc papAg, - 4 FIGURE 6.32 l t
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_, , . . ,., .,. - . . . _ - - - . - . . . , _ _ _ ~. . . . _ _ . . __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _
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j 1.525 - l 1.520 l 1.515 - ! 1.510 I
'1.505
'1.500 g 1.495 1.490 j i
f 1.485 : ] 1.480 1.475 i ., i IOP GAP BETWEEN RACKS 3 and 4 Rock COF=.5,all rocks ' i .4s5 . . . . "' 9 . . . .y. . . 3. . . 4. . . .g. . . g. . . .y. . . . , , , , . . . i'd " i'i i'i " i'i" i'4 " i's " i's TIME (sec.) FICURE 6.34
1 1.515 - i : . 1.510 { i i.50s ji 5 1.490 - i.485 -
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IL I i l l f [ TOP GAP BETWEEN RACKS 2 and 3 Rock COF=.5.oll rocks j i.4so
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! 0 : D 1.498 i i n : ! U 1.495 i : 1.493 i i l 1.490 i . 1 ~ 1.488 2: 2 FOP GAP BETWEEN RACKS 4 and 5 Rock COF=.s.oM rocks i 1.48s
- i O........d...5.......5.d 4
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l i I l I 7.0 ACCIDENT ANALYS13 and TERNAL (SECONDARY) STRESSES 7.1 Intrndnetian ' The James A. FitzPatrick Cafety Analysis Report has presented I results of analyses of several types of accidents which could ! potentially affect the spent fuel storage pools. Installation of , the proposed high density racks will enable the Authority to store increased amounts of spent fuel in the FitzFatrick plant spent fuel pool. Accordingly, accidents involving the spent fuel pool i have been reevaluated to ensure that the proposed spent fuel pool - modification does not change the present degree of assurance to public health and safety. The fallowing accidents have been l considered: 0 Fuel Pool - Earthquake Loading Loss of Water [ 0 Fuel Storage *=41 diag - Earthquake Loading I O Refueling Accidents - Dropped Fuel i Dropped Gates O Rack drop 7.2 n== nits af haa4daat maevainnelan ' l 7.2.1 Fuel Pool " The offeats of earthquake loadings on the fuel racks and spent fuel pool floor are discussed in sections 6.0 and . 9. 0 respectively of this report. The loss of cooling water in the spent fuel pool is discussed in section 5 of this report. G 7-1
7.2.2 Fuel Stormen Buildina The ability of the reactor building to. resist earthquakes,.has not been affected by ..the spent fuel pool d'> m f.ication. Therefore, the information presently contained in the FSAR is still valid. l 7.2.3 Refueling Accidents e This section considers three (3) accidents associated with the handling of fuel assemblies, the movement of transfer and reactor canal gates and the use of a fuel channel measuring , l-device. No other objects are ever moved over spent fuel. 7.2.3.1 Dreened Fuel Ama==hiv 3 The consequences of dropping a new or spent fuel assembly as it is being moved over stored fuel .is discussed below.
- a. Dropped Fuel Aasenhly Ameident I A fuel assembly is dropped from 24" a'bove a stcrage l location and impacts the base of the module. Local t
failure of the baseplate is acceptable; . however, . , the rack design should ensure that gross structural y failure does- not occur and the suberiticality of L the adjacent fuel assemblies is not violated. i calculated results show that the fuel assembly will l '- not hit the liner and that there will be no change in the spacing between fuel tubes. It is also shown that the load transmitted to the- liner through the support is well below that caused by seismic loads. If local deformation of the baseplate occurs, it is demonstrated that the liner-is not impacted. i L L l l 7-2 L l l l
.,. . , . . .- - . . ~ . . . - . - - . . . . . - - - . . _ . . - - - - . ~ . . - . . . . . . - - . - - .
i e J 1 i 1 l l 1
- b. Dronned Fuel Asammhly Aceldent II one fuel assembly dropping from 24" above the rack and hitting, the top of the rack. Permanent ,
defornation of the rack is acceptable, but is required to be limited to the top region such that { the rack cross-sectional geometry at the level of' q the top of the active fuel (and below) - is not altered. Analysis dictates that the maximum-local stress at the top of the rack is less than material. ' yield point. Thus, the functionality of the rack is not affected. If local deformation occurs, it is confined to a region above the active fuel area. '
- c. Dronned Fual Ama==hly Accident II l
This postulated accident is identical to (a) 'above - i except that the fuel assembly is assumed to drop in an inclined manner en top of the rack. Analyses show that the straight drop case (came b above) bounds the results. 7.2.3.2 Dropped Gate The reactor canal to pool gate is conservatively assumed-to fall from an elevation of 2 feet above the rack module. The gate is constructed of-aluminum, and weighs 370 lbs. in air. Its min 4=n= frontal areas corresponds , M an upright vertical fall. The mathematical model constructed to determine the impact velocity of the above falling object is based on several conservative assumptions, such as-
- a. The virtual mass of the body is conservatively assumed to be equal to its displaced fluid mass.
Evidence in the literature [1)' indicates t. hat the virtual mass can be many times higher.
- b. The mini === frontal area is used for evaluating drag coefficient.
7-3
j I
- c. The drag - coe'fficient . utilized in the analysis are t lower bound values reported in the literature (2].
In particular, at the beginning. of the fall when the velocity of the body is small, the Wrrasponding Re'enolds number is low resulting in a i larga drag coeff;.cient.
- d. The falling bodies are assumed to be rigid fori the purposes of impact stress calculation on the rack.- .;
The soldtion of the body motion problem is found. i analytically. The impact velocity thus computed is used. # to determine the =w4== stress generated due to stress wave propagation. 7.2.4 Rack._ Drop ' The scenario of a construction accident leading to a L rack dropping in the pool has been considered. It has been determined that a rack drop on an existing rack resulting in damage of stored fuel assemblies is not a credible scenario. The reasons for this conclusion are provided below:- A remotely engagable lift rig, meeting EUREG-0612 stress criteria, will be used to lift the empty modules. The - building - crane will be used for this purpose. A module installation scheme has been developed which ensures that all modules being handled are empty, and at least four feet laterally from a loaded module, when the module is more than six inches above the pool floor. Pursuant to the defense-in-depth approach of NUREG-0612, the fallowing additional measures of safety will be undertaken for the raracking operation. 7-4
t
'j (1) The crane and hoist will be given a preventive maintenance checkup and inspection within 3 months of the beginning of the.ruracking-operation. 1 r
(ii) The crane will be used to lift no' more than 504 of- ; its rated capacity at any time during the raracking operation. ' (iii)
'The rate of vertical movement will' not exceed fest per minute. 4>
(iv) The feet rate of horizontal movement will not exceed'5 ' per minute. " (v)
' Safe load. paths have been developed. The "old" or !
"new" racks will not be carried over any region of' the. pool containing fuel. i*
(vi) The rack upending. or laying down viil be carried' ' out in an area which is not proximate to any' safety related component. (vii) The installation crew will be given a minimum of four hours training in using the lift ' rig by.' the-rig designer. Video tapes of ' the rig showing its use, and application will be utilisW to train the ' crew in the proper use of the installation rig.
.The case of a heavy load dropping on the pool liner has been previously considered in the JAF FSAR, and this racking operation is covered by the previous safety evaluation in this matter.
7.3 7 m_t WTMG OF FmEt Eftt WatTE This sub-section and the . next one presents details on
.f the secondary stresses produced by bn@14ag and by temperature ;
effects. ' i; 7-5 _ - ,-e
1 t 1 The allowable local buckling stresses in the fuel cell , 1 walls are obtained by using classical plate buckling analysis. i The following formula for the critical stress has been used.
# rr2 Et2 12 b2 (1 p2) ,
where E = 27 x 106 psi, y is Poison's ratio, t= .075", b = 6.0". The factor A is suggested in (3) to be 4.0 for.a long panel loaded as shown in Figure 7.1. For the'given data acr < 15250 psi
- It should be noted that this calculation is based on the applied stress being uniform along the entire length of the cell wall. In ,
the actual fuel rack, the compressive stress' comes from consideration of overall bending of - the rack structures during a seismic event and as such is negligible at the rack ~ top and
==w4=n= at the rack bottom. It is conservative to apply the above equation to the rack cell wall if we compare acr with the maximum compressive stress anywnere in the cell wall. As shown in Section 6, this local buckling stress limit is not violated anywhere in the body of the ~ rack modules, since the ==w4=== compressive stress in the_ outermost cell-is a = 13890
- R6 (from. Table 6.5 with R6 =
.229) = 3181 poi.
7.4 ANALYSIS.OF WELDED.lDIHTS IN RACK $
^
, In-rack welded joints are awa=4n=4 under the icading conditions arising frca thamal effects due to an isolated het shell, in this sub-section. 7-6 graees r+=er=-. arm =e=** as oce==e+== *a e a e***- **-m-- *==mw me e g. =* * = = = + .e ,m. = ,
- 1 1
l A . thermal gradient between cells will develop when an isolated
-storage location contains a fuel assembly emitting maximum postulated heat, while the~ surrounding locations are-empty. We .l 1
can obtain a conservative estimate of weld stresses along - the length of_ an isolated hot cell by considering a beam strip j uniformly heated by 40*F, and restrained from growth along one ! long edge. The configuration'is shown in Figure 7.1. Using a shear beam theory, and subjecting the strip to a uniform temperature rise AT = 40*F, we can calculate an estimate of - the
=mvinne value of the average shear stress in the strip. The strip is subjected to the following boundary conditions.
- 1. -a. Displacement Us (x,y) = 0 at x = 0, at y = w/2, all x.
- b. Average force Mx, acting on the cross section Et = 0 at x = L, all y.
The final result for wall shear stress, ==vinn= at x = 1, is found to be given as EaAT raar "
.931 where E = 28 x 106 psi, a = 9.5 x 10-6 in/in *F and AT = 40*F.
Therefore, we obtain an estimate of maximum weld shear stress in - an isolated hot cell, due to thermal gradient, as i raax = 11550 psi since this is a secondary thermal stress, we use the allowable-l shear stress criteria for faulted conditions as a guide (t <
.42su )-
7-7
7.5 REFERENCES
[1] " Standards of Tubular Exchanger Manufacturer's Association", . 6th Edition, Section 12 (1978). [2] ." Fluid Machanies", by M.C. Potter and J.F. Foss, Ronald (1975), p. 454.
-[3) " Strength of Materials", - 5.P. Timoshenko, 3rd Edition, Part II, pp 194-197 (1956).
e se - 4 7-8
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. . ..itii . g-8.0 IN-EnvTcE suxvninMcE MtochM 8.1 PURPQSZ This section describes the programmatic commitments made by the New York Power Authority for in-service surveillance of the neutron absorption material (Boral). A poison surveillance program is presented in this section which allows access to representative poison samples without disrupting the integrity of the storage system. ~This program provides the capability to evaluate the poison material in a normal use mode and to forecast future changes. coupon surveillance: This procedure consists of preparing poison coupons carefully encased in a stainless steel metal jacket, and suspaaMag them from a " coupon tree". The ' tree" is suspended in a storage location selected in such a manner that it is surrounded by spent fuel discharged from the pool. 8.2 CXXTPQN SM7EIIL&MS 8.:.1 Descrintian_nf_Tast f.annons The poison used in the surveillance program will be representative of the material used within the storage system. It must be of the same composition, produced by the same method, and certified to the same. criteria as the. production lot poison. The sample coupon will be the same thieknaam as the poison used within 8-1 -
- ~
( l- . the storage system and will meet the referenced drawing dimensional requirements. Each poison specimen will be encased in a stainless steel jacket of an alloy identical to that used in~ the storage system, formed so as to encase the poison material and fix it in a position and with tolerances .similar to that for the storage racks. The jacket will be closed by quick disconnect clamps or screws with lock nuts in such a manner as to retain its form throughout the use period yet allow rapid and easy opening without contributing mechanical damage to the- poison specimen contained therein. A total of ten jacketed poison specimens will be used. The specimen location must be adjacent to a designated storage cell with design ability to allow for removal of the strip, providing access to a particular specimen. 8.2.2 Ranchmark_pata Benchmark tests will be performed on test coupons prior to their use. 8.2.3 Lena Ta m surveillaaan coupons viii be removed at scheduled intervals, and is awaminarl for loss of its physical and neatronic properties. G t* ;*
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l 9.O POOL STRUCTURAL ANALYETE
-Analyses have been performed to demonstrate that the increased leading on the pool floor and walls due to the re- 4 racking and subsequent accommodation of additional spent fuel '
assemblies can be accommodated in the JAF spent fuel pool without any modifications to the . pool or support columns. A non-linear limit strength analysis of the concrete-steel pool structure was '
- used to obtain the load carrying strength of the structure. The- ;
method of analysis was based on incrementing the applied loading ) L in steps and calculating the re-distribution of l'oad paths as each local portion of the structure reaches its permitted capacity. ' t, l The ANSYS finite element computer code was used to model the. spent' fuel pool along with the surrounding structure. The finite element model was used in conjunction. with the step by step loading procedure to track the progressive 194a? of the floor slab. In this manner, the load deflection curve of any point on the slab up to the slab limit load was determined by this model. All load combinations specified in NUR2G-0800 were considered. The controlling load combination, among.all load combinations considered, was found to be: ' 1.4 (Desf loat + Bydrostatic load) + 1.7 (Fuel Rack Dead Load) + 1.9 (==i==4 e load CBE) 4 9-1
9 The above combination can be found in Section 9.2 of ACI-349-85, modified per section.3.8.3. of'the USNRC-Standard Review Plan, Revision 1, July, 1981. Live load factor 1.7 was used ' for the fuel rack weight te provide additional conservatism in the analysis. The mathematical model describes the: fuel storage pool at elevation 3308-9", the East Wall which extends ' from the fuel _ pool floor to elevation 369 '-6", and both the north and south walls - which extend from elevation 311'-0" to 3698-6". In addition, the rooms west of the spent fuel pool from elevation 311'-0" to 369'- 6" were also modelled. The west wall (the reactor wall) ._ is -not included in the analysis; rather, it is assumed that this wall provides a fixed support to all attached structure. Figure 9.1 gives an overall view of the finite element model and Figure 9.2 shows the modelling of the floor slab. The floor slab is modelled using the triangular plate element from ANSYS in sufficiently fine detail to accommodate a non-linear analysis. In regions where fine detail is not required,-the ANSYS four noded rectangular plate element is used. The mathematical model contains 1615 plate elements and over 1400 nodo points. In addition, the spent fuel pool support columns were modelled as vertical. springs. The spring constant for the columns was adjusted as the loading increases to reflect the bottoaing out of the Belleville springs. 9-2
The section rigidity of the structure was calculated based on 9 bi-linear moment curvature relation. The flexural rigidity for , each range of behavior is the. slope of the moment curvature relation. The. steel reinforcing is assumed to be elastic-perfectly plastic, while the concrete structural behavior is ;
' described by well established concrete stress strain laws.. 1 Since the - ANSYS model is based on the total thickness, flexural rigidities are affected by the Young's Modulus assumed in each direction. While a section remains in the elastic range, the modulus is based on the transformed section inertia based on the appropriate concrete and steel layout. When yielding of the local section occurs, subsequent rigidity of the local section uses . a reduced modulus equal to 1% of the elastic modulus at the location. This permits the yielded zone to be characterized properly.
Thermal gradient effects in the floor slab were treated conservatively in the following manner. Initially, the concrete is assumed to be uncracked, reinforced, and fully restrained. This gives the largest t bending acaent across the slab due to a thermal gradient. This W 4 aa acaent results in a ==v4an= factored tensile thermal stress in - the bottom reinforcing of 10,790 psi at certain critical locations. This upper bound stress is very conservative as it is calculated neglecting all factors which would tend to- decrease its value. However, in order to conservatively treat the offact of thermal gradient, the slab sections subjected to positive acaents (tension in the bottom reinforcing bars) are assumed to be limited by yield stresses which are reduced by the conservatively calculated thermal stress. 9-3 e -- w --- - =- v,- r.-,,.
L The application - of all of the factored dead loads plus the hydtrostatic load is found not to yield any portion of- the ; structure. A series of incremental runs, using a unit pressure loading, is then carried out to track the subsequent behavior of < the system under the influence of the seismic load addition. . All required load combinations are considered, but the critical = loads combination is that given previously. Adjustments to the finita > elenant model are made at each step to reflect the strength redistribution eccurring as local portions of the structure reaches its limit' capacity. Poci slab analysis preceded the rack design, and therefore , utilized conservative input data. Table 9.1 gives the comparison between the data assumed in the analysis and the actual value. The results of the analysis showed that the allowable seismic load based on the slab's flexural capacity is 7.63 K5f. However, based on the beam shear capacity of the slab, the allowable seismic load capacity is only 2.75 E5f. Hence the capacity of the slab is controlled by shear. This 1W1= limit translates to a vertical ~ limiting peak OBE acceleration of .459 This is in aw maa of the .33g CBE peak level mandated for the plant so that - structural integrity is assured. For other load combinations, margins are obtained on allowable seismic load (either caE or SSE) which are greater than 1.36. The detailed rack dynamic analyses show that under SSE conditions, the added peak SSE load imposed on the floor, due to rack action, is bounded by the static load caused by a .2g vertical acceleration. 9-4
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.1 The above synopsis of the pool slab analysis ' indicates that sufficient load bearing capacity exists for the storage densification sought in this licensing application. _;
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7: Y. i 10.0 Padioloaical Consideratiens The additional spent fuel storage racks will not increase personnel radiation exposure during normal and refueling operations. Storing additional spent fuel'in the pool will increase the amount of corrosion and fission product radionuclides introduced into the pool water. Specifically, activated corrosion products ' such as Co-58, Co-60, Fe-59 and Mn-54 may be released to the pool water from the surface of the spent fuel assemblies and fission _ , products such as Cs-134, Cs-137, Sr-89 and Sr-90 may be released to the pool water through defects in the spent fuel cladding. However, the additional activity introduced to the fuel pool from the increase in stored fuel assemblies is not expected to increase radiation dose rates above the fuel pool which currently averages approximately 10mr/hr. Increasing the capacity of spent fuel assemblies near the walls j of the pool will not significantly affect the dose rates in the
- adjacent occupied areas, since they are adequately shielded by i the present 4 1/2 foot concrete walls surrounding the pool. This ,
shielding has been demonstrated by the fact that spent fuel has been stored close to the south and north walls with no adverse , consequences. 10.1 Solid PMwast;e It is estimated that no additional resins will be required by Spent Fuel Fool purification system during the rack addition . cork and therefore no additional resin shipments or increase in solid radioactive waste volumes will result due to rack addition. ! 10.2 Gaseous Releases No gaseous releases of radioactivity are expected as a result of adding racks to the Spent Fuel Pool. 10.3 Personnel Excesure Spent fuel which will eventually be stored in the new racks will be so well shielded by the water above the fuel (approximately 24') that radiation dose rates at the top of the pool will be i negligible, moreover, no routine activities occur on the outside ; of east side of the pool. Radiation surveys will be performed after spent fuel is placed in the new racks to verify the previous statements. 1 10-1 l l l l
10.4 Pediation Protection Durina Addino Packs ! 10.4.1 General Description of Protective Measures- t The radiation protection aspects of the spent fuel pool modification are the responsibility of FitzPatrick Radiclogical' and Environmental Services with the support of corporate staf f. Radiation, contamination and airborne surveys are performed prior L to any work in the pool and radiological conditions along with protective clothing requirements will be stated on applicable Radiation Work Permits. 10.4.2 Anticipated Exposures During Rack Addition . The effective dose rate in the work area of the Spent Fuel Pool is 3 mr/hr. The addition of five new racks is-estimated to take approximately two weeks with.6 workers and 2 radiation protection technicians. Based on these parameters, the total estimated exposure for the project is 2 person rem. l 10-2 1' l l l l 1
. i a
o f t f-11.1 COST / BENEFIT ASSESSMENT A cost / benefit assessment by the Authority demonstrated that the installation of additional high density spent fue.t a storage racks is the most advantageous means of providing necessary spent fuel storage considering public safety and i projected costs. 11.1 NEED FOR INCREASED STORAGE CAPACITY A. The Authority currently has no contractual arrangements with any fuel reprocessing' facility. There are no operating or planned fuel reprocessing facilities available in the U.S. , The-Authcrity has executed contracts with the U.S. t Department of Energy (DOE) pursuant to the Nuclear Waste L Policy Act of 1982. However, the disposal facilities arc not expected to be available for spent fuel any earlier than 2003 (Reference 1). B. Table 11.1 includes a projected refueling schedule for J.A. PitzPatrick and-the expected number of fuel asse blies that-will be transferred into spent fuel pool at each refueling until the ability to maintain a full core reserve is lost in 1991. At present, the licensed capacity is 2244 storage cells. All calculations in the table for loss of full core 11-1 i l
1 e t reserve (FCR) are based on the nur.ber of licensed total
-calls in the pool. The table in then continued assuming the installation of 553 additional cells which lengthens the time of loss of FCR to the year 1997.
- c. Adoption of this proposed spent fuel storage expansion would not necessarily extend the time period that spent fuel assemblies would be stored-en site. Spent fuel will- :
be removed from the site for disposal under the provisions-of the Nuclear Waste Policy Act of 1982, but a government facility is not currently expected to be available to accept full reload quantities of spent fuel from J. A. FitzPatrick before 2005-(Reference 2). t 11.2 ESTIMATED COSTS
' Total construction cost associated with the-proposed modification is estimated to be approximately 4.1 million del?ars. This figure includes the cost of designing and f6hricating the spent fuel racks; engineering costs;- and installation and support costs at the site. -
11.3 CONSIDERATION OF. ALTERNATIVES A.- There are no operational commercial reprocessing facilities available for Attthority's needs in the United States, nor are there expected to be any in the foreseeable future. E. While plans are being for=ulated by DOE for construction of spent fuel disposal facilities per the Nuclear Kaste Policy Act of 1952, a facility is net expected to be available to 11-2
l accept spent fuel any earlier than 2003 (Reference 1). C. The Authority does not own or control any facility where it could transfer spent fuel from J.A. FittPatrick. The Indian Point.3 nuclear plant, owned by the Authority, is a Pressurized Water Reactor (PWR) with PWR spent fuel racks that are not designed to accept Boiling Watar Reactor (BWR) fuel from J.A. FitzPatrick. D. There are no existing available independent spent fuel . storage facilities. Transfer of J.A. FitzPatrick spent fuel to other utility facilities would only compound storage problems there and is not a viable option. E. Licensed at-reactor spent fuel storage alternatives involving dry cask / vault storage were evaluated and excluded from consideration at this time due to overall economic reasons. These alternatives, as well as other technologies such as fuel consolidation, would be further evaluated-if a further need to expand the spent fuel storage capability at J.A. FitzPatrick becomes necessary. F. Estimates for costs of replacement power were' calculated in Table 11.2 based on the New York Public Service commission's avoided capacity and energy costs as per cases no. 28962, 28793 and 28689 dated May 10, 1988. Annual and cc=ulative replacement power costs are given starting in' 1998, the first year spent fuel in the reactor could not be removed due to lack of storage capacity in the existing 11-3 l
l 1 l racks, through the yesr 2002. This scenario anticipates that the U.S. Department of Energy will be removing fuel from-J.A. FitzPatrick at a rate equal to the generation rate by the year 2003. I J.A.lFitzPatrick power is now used by six of the seven New York State Investor-owned Utilities, Municipal Cooperative and Industrial customers. Plant shutdown would place a-heavy financial burden on New York residents served j by the Authority and cannot be justified. 11,4 RESOURCES COMMITTED t The rack addition to the spent fuel pool will.not result in any irreversible and irretrievable commitments of water, land, and air resources. The land area now used for the spent fuel. pool will be used more efficiently by safely increasing the total number of available storage cells. The materials used for new rack fabrication are discussed in Section 3.3. These materials are not expected to significantly foreclose alternatives available with respect to any other licensing actions designed.to improve . L the capacity for' storage of spent fuel. c 11.5 TEERMAL IXPACT ON TEE ENVIRO 10 CENT Section 5.0 censidered the following: the additional l heat load and the anticipated maximum te perature of water L , in the spent fuel pool that would result fr== the preposed expansion, the additional heat load en cc p:nent and/or 1 1. 11-4 l l
)
l i i
.. 1 plant cooling water systems, the bulk peel-and' local water temperatures as well as the time-to-boil if plant cooling .
. systems'should become inoperative, and whether there will fl be any significant increase in the amount of heat released to the environment. The proposed increase in storage capacity will result in an. insignificant impact on the i-environment.
11.6- REFERENCES -
- 1. U.S. Department of Energy, " Draft 1988 Mission Plan Amendment," June 1988.
- 2. U.S. Department of Energy, " Annual Capacity Report,"
June 1988. l
.I i
i 1 11-5 l l y
TABLE 11 1-NUCLEAR FUEL DISCHARGE INFORMATION J.A. FITZPATRICK Number Cumulative Total of Cycle Shutdown Assemblies Spent Fuel Assemblies-No. Dates Discharge in the Pool 01 6/1977 132 132 02 9/1978 136 268 160 428 03 5/1980 616 04 11/1981 188 05 6/1983 200 816 06 2/1985 196 1012 07 1/1987 188 1200 08 8/1988 184 1384 2244 Currently Installed cells ( ACAL CYCLE INFOPF.ATION TEROUGH CYCLE 08, PROJECTED TFIPIAFTER) 09 3/1990 156 1540(3) 10 10/1991 208 1748 11 - 10/1993 204 1952 12 10/1995 208 2160(2) 13 10/1997 208 2368 14 10/1999 208 2576 15 10/2001 208 2784 16 10/2003 208 2992 17 10/2005 208 3200 18 10/2007 208 3408 19 10/2009 208 3616 20 10/2011 208 3824 21 10/2013 208 4032 END OF LIFE 10/2015 560 FINAL OFFLOAD 4592 (1) TULL COP 2 RESERVE (FCR) LOST AT 16S4 CELLS WITH CUPSINT RACYJ; FACK ADDITION F22UIPID TO ?.I2AIN FCR (2) FCR,M ST AT 2237 CELLS WITH RACK AIDITION (2797 AVAIIAELE STORAGE w w wnas 11-6 _ - _ _ - _ - - _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ [
-3 TABLE 11.2 ANNUAL REPLACEMENT POWER COSTS ATTRIBUTED TO J.A. FIT 2 PATRICK Cumulative Nominal Net Cumulative Present Value Present value Energy 1989 Dollars 1*rodugon Replacement Costa (2) Cost 1989Dogra ($000)
Year (GWII) ($000)- ($/MWil) ($000) ($000) 89.20 486,417 243,330 243,330 1998 5,153 486,417 478,738 5,453 508,229 93.20 994,645 235,408 1999 1,525,232 227,569 706,307 2000 5,453 530,587 97.30 926,105 553,490 101.50 2,078,722 219,798 2001 5,453 221,763 1,147,868 2002 5,153 603,113 110.60 2,681,835 (1) Based on: Plant rating of 830 MW and annual capacity factor of 75%. (2) Calculated based on statewide avoided capacity and energy costs in the Niagara Mohawk franchise area prepared and issued by the New York Public Service Commission on 5/10/88 as per Cases No. 28962, 23793 and 28689. Reflects gross replacement costs (excludes any offset for avoided variable costs such as fuel and operation and maintenance expenses). l (3) Dased on a discount rate of 8%. NOTE: J. A. FitzPatrick assumed to be out of service all year (s). 11-7 1 l
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